ML14094A155
| ML14094A155 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/25/2014 |
| From: | NRC Region 1 |
| To: | Entergy Nuclear Generation Co |
| Jackson D | |
| References | |
| TAC U01876 | |
| Download: ML14094A155 (223) | |
Text
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 300000 K1.02 Importance Rating 2.7 Knowledge of the connections and I or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: Service air Proposed Question: RO Question # 1 Given the following:
- Low pressure service air is being supplied via the HI/LOW Service Air cross connect
- An air leak occurs at some unknown location
- The K-117 Air Compressor starts and stabilizes air system pressure at 78 psi.
Assuming that NO operator action has been taken:
(1) What is the position of A0-4350, Serv Air Hdr Block Valve AND (2) A0-4353, High Press/Low Press Service Air Inlet Crosstie Valve?
A0-4350 1 Serv Air Hdr A0-4353 1 High Press/Low Press Block Valve Service Air Inlet Crosstie Valve A. Open Open B. Open Closed C. Closed Open D. Closed Closed Proposed Answer: D Explanation (Optional):
A. Incorrect: A0-4353 closes at 85 psig. Plausible in that A0-4365, Non-Ess lnstr Air Hdr Blk Vlv, closes at 80 psig.
B. Incorrect: A0-4350 closes at 85 psig. Plausible in that A0-4365, Non-Ess lnstr Air Hdr Blk Vlv, closes at 80 psig.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: Both valves close at 85 psig. Plausible in that A0-4365, Non-Ess lnstr Air Hdr Blk Vlv, closes at 80 psig.
D. Correct: Per PNPS 5.3.8, at 85 psig A0-4350, Serv Air Hdr Block Vlv, closes to isolate the service air header and A0-4353, High Press/Low Press Service Air Inlet Crosstie Valve, closes to isolate instrument air from low pressure service air header.
Technical Reference(s): PNPS 5.3.8, pg 3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-02-04, E0-9 (As available)
Question Source: Bank # WTSI Bank 12738 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 262002 K1.19 Importance Rating 2.9 Knowledge of the physical connections and/or cause- effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following: Power range neutron monitoring system: Plant-Specific Proposed Question: RO Question # 2 A loss of 120 VAC Vital Bus Y-2 has occurred. Which one of the following is correct regarding the ability to monitor reactor power on Panel C905?
A. APRM and IRM recorder indications are lost APRM, IRM and SRM status lights are lost B. APRM and IRM recorder indications remain functional APRM, IRM and SRM status lights remain functional C. APRM and IRM recorder indications are lost APRM, IRM and SRM status lights remain functional D. APRM and IRM recorder indications remain functional APRM, IRM and SRM status lights are lost Proposed Answer: C Explanation (Optional):
A. Incorrect: APRM, IRM and SRM status lights are not lost. Plausible in that these are lost if a loss of 120 VAC Bus Y-1 has occurred.
B. Incorrect: APRM and IRM recorder indications are lost. Plausible in that they will remain functioning if a loss of Instrument bus Y-1 occurs.
C. Correct: APRMIIRM recorder indication is lost.
Also lAW section 5.0 PNPS 5.3.6, Discussion Section item 1, the indications below are still available on C905 panel:
- All SRM/IRM/APRM/RBM status lights on the C905 horizontal section PNPS 2014 NRC Written 022014.doc
D. Incorrect: APRM and IRM recorder indications are lost. Plausible in that they will remain functioning if a loss of Instrument bus Y-1 occurs.
Technical Reference(s): PNPS 5.3.7, Discussion Section 4 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-01-07, EO 5b (As available)
Question Source: Bank#
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 203000 K2.02 Importance Rating 2.5 Knowledge of electrical power supplies to the following: Valves (RHRILPCI: Injection Mode)
Proposed Question: RO Question # 3 The 480 VAC distribution is aligned for normal full power operation when 480 VAC Load Center B-1 trips.
The status of the B-1 to B-6 supply breakers is as follows:
- Breaker 52-102 on B-1 is closed (B-6 Supply Breaker)
- Breaker 52-601 on B-6 is open (B-6 Supply Breaker)
What is the status of power availability to the following valves?
- LPCIInjection Valves #2, M0-1001-29A and B
- LPCIInjection Valves #1, M0-1001-28A and B
- Recirc Pump Discharge Valves, M0-202-5A and B A All valves have power B. NO valves have power C. M0-1001-298 has power M0-1001-28B has power M0-202-5B has power All other valves do NOT have power D. M0-1001-29A and B have power M0-1001-28A and B have power M0-202-5B has power M0-202-5A does NOT have power Proposed Answer: B Explanation (Optional):
A Incorrect: Bus B-6 is de-energized. Swing bus B-6 will not transfer to B-2 unless both PNPS 2014 NRC Written 022014.doc
breakers from B-1 are open. Bus B-6 supplies all valves that must re-position on a LPCI loop selection via B-20. All of the noted valves may be required to reposition and are therefore are without power. Plausible if the candidate thinks that with the one supply breaker open, B-6 will transfer.
B. Correct: 480 V LC B-6 can be energized from B-1 (normal) or B-2 (alternate). Normally if bus B-1 suffers loss of voltage for > one second,52-102 and 52-601 will trip open and breakers52-202 and 52-602 automatically close. However if the B-1 supply,52-102 fails to open the corresponding supply from B-2,52-202 will be interlocked open.
Therefore B-6 will not auto transfer. Bus B-6 supplies all valves that must re-position on a LPCI loop selection via B-20. All of the noted valves may be required to reposition and are therefore without power.
C. Incorrect: All valves are without power. Plausible if the candidate thinks the "B" side valves are powered from the "B" side of the electric plant and not the swing bus.
D. Incorrect: All valves are without power. Plausible if the candidate does not realize that the Recirc valves that must reposition on a LPCI loop select are also powered from swing bus B-6 and concludes that the "B" Recirc discharge valve is still has power available.
Technical Reference(s): 480 VAC Reference Text, pages (Attach if not previously provided) 16, 39, 40, 41 and Figure 2 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-01, EO 15a (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 PNPS 2014 NRC Written 022014.doc
Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215003 K2.01 Importance Rating 2.5 Knowledge of electrical power supplies to the following: IRM channels/detectors Proposed Question: RO Question # 4 IRM Channel "C" has just gone INOP.
Which one of the following electrical losses could have resulted in this occurrence?
A. 24 VDC D-25 trip B. 24 VDC D-26 trip C. 120 VAC Instrument Bus Y-1 trip D. 120 VAC Safeguard Power Supply Y-3 trip Proposed Answer: A Explanation (Optional):
A. Correct: D-25 supplies IRM Channels A, C, E, and G. An INOP trip is generated on loss of high voltage power supply.
B. Incorrect: D-26 supplies IRM Channels B, D, F, and H.
C. Incorrect: Y-1 supplies the detector drive.
D. Incorrect: Y-3 does not supply any component within the IRM system.
Technical Reference(s): 24 VDC Distribution Reference (Attach if not previously provided)
Text, page 7 IRM Reference Text, page 18 5.3. 7, Discussion, [5]
Proposed References to be provided to applicants during examination: None PNPS 2014 NRC Written 022014.doc
Learning Objective: (As available)
Question Source: Bank #
Modified Bank# WTSI 1274 Modified for Pilgrim and changed stem to say INOP trip vice just loss of power.
New Question History: Last NRC Exam: 2007 Fermi Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 211000 K3.01 Importance Rating 4.3 Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on following: Ability to shutdown the reactor in certain conditions.
Proposed Question: RO Question# 5 During an ATWS, Standby Liquid Control (SLC) injection is required. The following sequence occurs:
- The operator places SLC ACTUATE switch to "SYS A"
- SLC Pump "A" starts
- The operator places SLC ACTUATE switch to "SYS B"
- SLC Pump "B" starts and injects
- The breaker for SLC Pump "B" trips 15 seconds later Which one of the following is correct regarding any required actions to shutdown the reactor using boron injection?
A. Boron can only be injected using RWCU.
B. The SLC ACTUATE switch should be switched to "SYS A".
C. The SLS System "A" local pushbutton must be depressed.
D. Attempt one reset of the SLC Pump "B" breaker. If the breaker trips again, inject with RWCU.
Proposed Answer: B Explanation (Optional):
A. Incorrect: Boron will inject if the SLC ACTUATE switch is placed to "A".
B. Correct: The discharge of the two pumps are cross-tied. Either pump will inject thru either squib valve. Although the "A" squib failed to fire, the "A" pump started. When SYS "B" was started the pump was injecting thru the "B" squib valve. After the "B" pump tripped, if the switch is placed back in SYS "A", the "A" pump will restart and inject thru the "B" squib.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: The local pushbutton will only start the pump; it will not fire the squib valve.
D. Incorrect: Injection capability is available from the control room.
Technical Reference(s): 2.2.24, Sect. 7.2 (Attach if not previously provided)
SLC Reference Text, Figure 1 SLC Reference Text, local controls Proposed References to be provided to applicants during examination:
Learning Objective: 0-R0-02-06-06, EO 10 (As available)
Question Source: Bank # PNPS 15625 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215005 K3.07 Importance Rating 3.2 Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: Rod Block Monitor Proposed Question: RO Question # 6 A power ascension with control rods is in progress. Current reactor power is 52% and a center control rod is being withdrawn. During the rod withdraw, APRM channei"B" fails downscale.
Assuming no operator actions, which one of the following describes the response of the Rod Block Monitoring System?
A. A RBM INOP trip is generated B. A RBM Downscale trip is generated C. An automatic bypass of RBM channel "B" occurs D. The reference APRM signal to RBM "B" automatically shifts to APRM "D" Proposed Answer: C Explanation (Optional):
A. Incorrect: The RBM automatically bypasses.. Plausible if the candidate knows that the RBM uses the APRMs and concludes that the reference APRM failure will cause a RBM lnop trip.
B. Incorrect: The RBM automatically bypasses. Plausible if the candidate recalls that the APRM is the reference APRM for RBM "A" and concludes that the reference APRM is used for determining that the RBM is downscale vice no longer required.
C. Correct: APRM "B" is the "Normal" reference to RBM "A". When the APRM fails downscale, the RBM responds as if power is< 26%. The RBM is not required below 26% and an automatic RBM bypass occurs.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: The RBM automatically bypasses. Plausible if the candidate recalls that there is a backup reference APRM for each RBM channel. However the RBM will not shift to its backup reference until the "Normal" APRM is bypassed. APRM "D" is the backup for RBM"B".
Technical Reference(s): RBM Reference Text, pages 11, (Attach if not previously provided) 12 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-07 -05, EO 13b (As available)
Question Source: Bank # 227 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: 2009 Pilgrim, Question 6 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 206000 K4.14 Importance Rating 3.4 I
Knowledge of HIGH PRESSURE COOLANT INJECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Control oil to turbine speed controls:
Proposed Question: RO Question # 7 1 HPCI operation is required to maintain adequate core cooling. However rated injection flow is not required.
Which one of the following is a system operating limit that ensures that there is adequate control oil pressure to the turbine speed controls?
Maintaining HPCI ....
A. Injection ~ 1000 gpm B. Injection~ 3000 gpm C. Turbine speed ~ 1000 rpm D. Turbine speed ~ 3000 rpm Proposed Answer: C Explanation (Optional):
A. Incorrect: The limit is 1000 rpm on the turbine. Plausible if the candidate confuses the RPM limit as an injection limit as there is a limit on gpm injection (see below.) However this limit applies to automatic flow control.
B. Incorrect: The limit is 1000 rpm on the turbine. Plausible in that this is the limit for operating the HPCI turbine in automatic flow control.
C. Correct: lAW PNPS 2.2.21.5, Attachment 1, Section 4.0 Caution, during normal operation, the HPCI turbine should not be run below 2000 RPM. Below 2000 RPM, intermittent exhaust flow will cause water hammer in the exhaust line. If HPCI turbine operation below 2000 RPM is required to achieve and/or maintain adequate core cooling, then the HPCI turbine should not be run below 1000 RPM. This will ensure adequate oil pressure to the control oil system and bearing lubrication.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: The limit is 1000 RPM. Plausible in that there is a limit associated with the number 3000 but it is associated with injection flow and has to do with controller operations.
Technical Reference(s): 2.2.21.5, Attachment 1, Section (Attach if not previously provided) 4.0 Caution Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-03, EO # 11 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 209001 K4.02 Importance Rating 3.0 Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: Prevents water hammer Proposed Question: RO Question # 8 Core Spray is aligned for normal full power operation.
Which one of the following is correct regarding the prevention of water hammer in the core spray system?
A. Condensate Transfer system pumps maintain both the Core Spray pump suction and discharge piping full of water.
B. Condensate Transfer system pumps maintain the Core Spray pump discharge piping full of water. Torus water maintains the Core Spray pump full.
C. Demin Water Transfer system pumps maintain both the Core Spray pump suction and discharge piping full of water.
D. Demin Water Transfer system pumps maintain the Core Spray pump discharge to keep the discharge piping full of water. Torus water maintains the Core Spray pump full.
Proposed Answer: B Explanation (Optional):
A. Incorrect: Torus water maintains the pump suction and pump piping full of water.
Plausible if the candidate believes that Condensate Transfer is supplied to the Core Spray pump suction. This is plausible if the candidate recalls that Condensate transfer water is supplied to the RHR pump suction line (for flushing purposes).
B. Correct: Water from the Condensate Transfer System pumps (Transfer or Jockey pumps) is supplied to the Core Spray discharge /injection line downstream of the pump discharge check valve. This "Keep-fill" maintains the piping pressure at 110 psig to prevent water hammer and piping damage when the CS pumps start.
Torus water maintains the pump suction and pump full of water up to the pump discharge valve. Condensate Transfer pump discharge pressure is much higher than PNPS 2014 NRC Written 022014.doc
the static pressure of the torus and the Core Spray pump discharge check valve is kept closed.
C. Incorrect: Condensate transfer supplies the water. Plausible in that Demin water provides makeup to various systems.
D. Incorrect: Condensate transfer supplies the water. Plausible in that Demin water provides makeup to various systems.
Technical Reference(s): Core Spray reference text, page (Attach if not previously provided) 14 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-02, EO 7d (As available)
Question Source: Bank #
Modified Bank # PNPS Bank# 2875 Modified stem to address core spray vice RHR. Also modified to require candidate to how each section of piping is maintained full of water.
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 259002 K5.01 Importance Rating 3.1 Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: GEMAC/Foxboro/Bailey controller operation: Plant-Specific Proposed Question: RO Question # 9 A reactor plant shutdown is in progress. The "B" Feed Reg Valve was placed in manual as power was lowered to steady out observed water level control oscillations.
Shortly after commencing the down power ...
- A Recirc pump trips.
- Reactor power stabilizes at 65%.
- RPV water level stabilizes at 30 inches
- FWLC individual M/A station's are currently indicating as shown in the picture to the right Given these indications and assuming the power reduction continues, which one of the following is correct regarding:
( 1) The response of the controller to the Recirc pump trip AND (2) Any require action?
A. (1) The "A" M/A station is NOT responding properly. If not corrected a low vessel level scram will occur.
(2) Place "A" FRV in manual, increase the output signal to balance flows.
B. (1) The "A" M/A station IS responding properly. If not corrected a turbine trip on high level will occur.
(2) Lower the output signal of the "B" FRV at the "B" M/A station.
C. (1) The "B" FRV is NOT responding properly. If not corrected a turbine trip on high PNPS 2014 NRC Written 022014.doc
level will occur.
(2) Lock the FRV in the condenser compartment and attempt to throttle the valve closed to balance flows D. (1) The "B" M/A station IS responding properly. If not corrected a low vessel level scram will occur.
(2) Raise the Setpoint on the Master Controller to open the "A" FRV further.
Proposed Answer: B Explanation (Optional):
A. Incorrect: The "A" controller is responding correctly in Auto. The "B" controller being in manual, maintained the "B" FRV at the position required for high power operation. After the Recirc pump tripped, level rose due to the amount of feed going thru the "B" Recirc pump. The "A" controller cut back the "A" FRV in response. Going to manual and raising the feed thru "A" FRV will only cause level to rise.
B. Correct: The "A" controller is responding correctly in Auto. The "B" controller being in manual, maintained the "B" FRV at the position required for high power operation. After the Recirc pump tripped, level rose due to the amount of feed going thru the "B" Recirc pump. The "A" controller cut back the "A" FRV in response. If the down power continues the "A" FRV will go full close. Any additional reduction will result in a rising level and turbine trip. The required action is to lower the flow thru the "B" FRV which will cause the "A" FRV to open further to maintain level providing more margin to the valve going close and will balance feed flows.
C. Incorrect: The "B" FRV is responding properly. Its output is higher than "A"'s only because it was in manual and set to maintain level at the initial high power condition.
Plausible if the candidate focuses on the output signal and concludes that the controller has failed. In that case the action described would be correct.
D. Incorrect: If the power reduction is continued, the "A" FRV will go close, level will begin to rise and eventually reach the high level turbine trip. Raising the Auto setpoint will only cause a momentary opening of the "A" FRV which will eventually return to its original position.
Technical Reference(s): FWLC Reference Text, page 18 & (Attach if not previously provided) 19 and figure 1a and 9 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-04-10, EO 13 (As available)
PNPS 2014 NRC Written 022014.doc
Question Source: Bank # LOR Bank# 43 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215004 K5.03 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM: Changing detector position Proposed Question: RO Question # 10 A reactor startup is in progress. The Reactor Mode Switch is in STARTUP and aiiiRMs are on range 2 SRM Channel "A" is selected for withdraw and the Withdraw pushbutton has just been depressed.
Which one of the following will be the FIRST to occur as the detector is withdrawn?
A. SRM Downscale rod block B. SRM Inoperable rod block C. Detector retract not permitted rod block D. Automatic termination of the detector withdraw Proposed Answer: C Explanation (Optional):
A. Incorrect: The SRM downscale will not occur until count rate drops to 3 CPS.
B. Incorrect: An SRM inoperable trip is caused by:
- Low detector voltage
- SRM module unplugged
- The SRM function switch is not in OPERATE C. Correct: If any SRM channel detects < 100 cps and its detector is not fully inserted in the core, a rod block is activated. This rod block is bypassed if:
- The mode switch is in RUN
- The associated IRM range switches are at range three or above (IRM channels A, C, E, and Grange switches bypass SRM A and C, IRM channels B, D, F, and PNPS 2014 NRC Written 022014.doc
H bypasses SRM Band D).
Since the IRMs are on range 2, this rod block will occur.
D. Incorrect: Detector will not automatically stop. Plausible if the candidate concludes that this action will occur to prevent the downscale rod block at 3 CPS.
Technical Reference(s): SRM Reference Text, pages 17 (Attach if not previously provided) and 18 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-07 -01, EO 6 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not Used I
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 264000 K6.08 Importance Rating 3.6 Knowledge of the effect that a loss or malfunction of the following will have on the EMERGENCY GENERATORS (DIESEL/JET): A.C. power Proposed Question: RO Question # 11 The startup transformer is supplying all 4160 VAC buses.
Then the following occurs:
- Due to problems on the 345 KV electrical system, voltage on Pilgrim's 4160 VAC distribution begins to lower.
- Bus undervoltage alarms for both A5 and A6 buses annunciate.
- Bus voltage drops to 381 0 volts and stabilizes at this value.
What automatic actions should take place for this occurrence?
Startup transformer breakers for A5 and A6 will ....
A. trip after approximately 1.2 seconds. Diesel ,generators will start and power A5 and A6.
B. trip after approximately 10 seconds. Diesel ~1enerators will start and power A5 and A6.
C. remain closed but load shed will initiate after approximately 10 seconds.
D. trip after approximately 10 seconds. Diesel generators will not start but A5 and A6 will be powered by the shutdown transformer.
Proposed Answer: B Explanation (Optional):
A. Incorrect: Breakers will not trip for 10 seconds. Plausible in that there is a 1.2 breaker trip but that will not occur unless voltage drops to < 3780 volts.
B. Correct. If the Startup Transformer feeder breaker to A5 or A6 is CLOSED, then the breaker will trip when Startup Transformer secondary voltage is less than or equal to 3858.92 to 3898.48V after a 9.88 to 10.60-second time delay. The diesels will start if:
PNPS 2014 NRC Written 022014.doc
- The Aux Trans breaker is open AND
- The Startup transformer (S/U Xfmr) breaker is open AND
- Degraded S/U Xfmr. Secondary Volts, 3880 volts for~ 10 seconds So when the S/U transformer breaker was tripped open by the degraded voltage condition, the diesels started and re-energized their buses.
C. Incorrect: Load shed will not initiate because an accident signal is not present.
Plausible in that the undervoltage alarms input into the load shed circuit and if the candidate believes that load shed is initiated in an attempt to recover voltage. However this will only occur if an ECCS start signal was also present.
D. Incorrect: The diesels will re-power the buses. Plausible if the candidate does not think the EDGs will start at this voltage level. If so, then the shutdown transformers would energize the buses.
Technical Reference(s): Emergency AC Reference test, (Attach if not previously provided) page 18 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-08 E0-4 (As available)
Question Source: Bank # 14153 Made editorial changes to stem and replaced distractor "A" Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 217000 K6.01 Importance Rating 3.4 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): Electrical power Proposed Question: RO Question # 12 The plant is at rated conditions with RCIC in its normal standby lineup.
Which of the following power losses will prevent RCIC from auto starting and injecting at rated flow if the initiation setpoint was reached?
- Loss of 480 VAC MCC B-18
- Loss of 125 VDC Bus D-6
- Loss of 125 VDC Bus D-16
- Loss of 125 VDC Bus D-17 A. loss of 125 VDC Bus D-16 only B. loss of 125 VDC Bus D-17 only C. loss of 125 VDC Bus D-6 or the loss of 480VAC MCC B-18 D. loss of 125 VDC Bus D-16 or the loss of 480VAC MCC B-18 Proposed Answer: A Explanation (Optional):
A. Correct: A loss of D-16 will prevent RCIC from starting as all motor operated valves and control circuitry except for M0-1301-16 from power center D-16. D-16 supplies D4 and D7. D4 supplies the RCIC inverter, and D-7 supplies the DC MOVs.
B. Incorrect: A loss of D-17 will not prevent RCIC from starting. Plausible in that it will prevent HPCI from starting.
C. Incorrect: Neither a loss of D-6 or B-18 will prevent RCIC from starting. Loss of B-18 will not prevent RCIC from auto starting. Plausible in that B-18 powers Steam Supply Valve M0-1301-16. However this valve is normally open when RCIC is in its standby lineup.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: Loss of B-18 will not prevent RCIC from auto starting. Plausible in that B-18 powers Steam Supply Valve M0-1301-16. However this valve is normally open when RCIC is in its standby lineup.
Technical Reference(s): PNPS 5.3.11 (Attach if not previously provided)
PNPS 5.3.12 125 VDC Reference Text, page 30 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-04, EO 13 (As available)
Question Source: Bank#
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 239002 A1.08 Importance Rating 3.8 Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including: Suppression pool water temperature Proposed Question: RO Question # 13 The plant is starting up from an outage that included replacing the SRVs.
Testing of the new SRV is in progress.
lAW Tech Specs which one of the following is the LOWEST Torus water temperature which requires that the SRV testing be secured?
Proposed Answer: B Explanation (Optional):
A. Incorrect: Testing can continue up to 90°F. Plausible in that 80°F is the maximum temperature during normal continuous operation lAW TS 3. 7 .A.1.c. However during testing the limit is raised to 90 degrees.
B. Correct: lAW TS 3. 7.A.1.d, the maximum suppression pool bulk temperature during RCIC, HPCI or ADS operation shall be s 90"F. TS 3.7.A.1.f goes on to say that If the suppression pool bulk temperature exceeds the limits of Specification 3.7.A.1 .d, RCIC, HPCI or ADS testing shall be terminated and suppression pool cooling shall be initiated.
C. Incorrect: Testing must be secured when torus temperature exceeds 90 degrees.
Plausible in that this is the TS limit at which a scram is required as required by TS 3.7.A.1.g.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: Testing must be secured when torus temperature exceeds 90 degrees.
Plausible in that this is the TS limit at which the vessel must be depressurized to s 200 psig as required by TS 3. 7.A.1.g.
Technical Reference( s): TS 3. 7 .A.1.c, d, f, g (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-03, EO 15g (As available)
Question Source: Bank#
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Shielding, isolation, and containment design features, including access limitations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 205000 A 1. 05 Importance Rating 3.4 Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including:
Reactor water level Proposed Question: RO Question # 14 Shutdown Cooling is being placed in service and the RHR pump is about to be started.
lAW PNPS 2.2.19.1, Shutdown Cooling Operations, which one of the following is correct regarding:
(1) the indicated RPV level response when the RHR pump is first started AND (2) the required action taken before starting the RHR pump in anticipation of that indicated level response?
A. ( 1) Level will lower (2) Raise level to > +60 inches B. ( 1) Level will lower (2) Raise level to > +30 inches C. ( 1) Level will rise (2) Lower level to between +20 and +25 inches D. (1) Level will rise (2) Align RWCU for reject operation Proposed Answer: B Explanation (Optional):
A. Incorrect: The required action is to raise level to> +30 inches. Plausible if the candidate confuses the water level requirement for a loss of SOC cooling (Raise RPV PNPS 2014 NRC Written 022014.doc
level to > +60 inches).
B. Correct: RPV level will lower when the pump is started. PNPS 2.2.19.1 directs that RPV level be raised to > +30 inches in order to prevent a low RPV level SOC isolation when the pump is started.
C. Incorrect: RPV level will lower when the pump is started. Plausible if the candidate confuses the plant response with what occurs when securing the RHR pump. If so the candidate would apply the required action prior to securing the pump (lower level to between +20 and +25 inches).
D. Incorrect: RPV level will lower when the pump is started. Plausible if the candidate confuses the plant response with what occurs when securing the RHR pump. Although plausible, the required action is not to verify or align RWCU for reject operation.
Technical Reference(s): PNPS 2.2.19.1, page 46 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-01, TO, 205-01-01-018 (As available)
PLACE RHR IN SHUTDOWN COOLING.
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 223002 A2.1 0 Importance Rating 3.9 Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of coolant accidents Proposed Question: RO Question# 15 The plant was at rated conditions when a loss of coolant accident occurred. Current plant conditions are as follows:
- The Reactor Mode Switch is in Shutdown
- Drywell pressure is 12 psig and rising
- RPV level is+ 58 inches and slowly rising due to swell
- RPV Pressure is 700 psig and slowly rising
- Main Steam Tunnel temperature is 160 oF and rising slowly due to a loss of ventilation
- HPCI injected but is now tripped on high level
- The MSIVs have automatically closed
- EOPs require that the MSIVs be reopened in order to perform an Emergency Depressurization using the bypass valves Which one of the following is correct regarding:
- The Group 1 MSIV isolations signals that ar,e currently present AND
- lAW PNPS 5.3.21, the actions directed to defeat the isolation signal(s) in order to reopen the MSIVs?
A. ( 1) High RPV Level only (2) The PCIS RX WTR LVL Bypass switches on Panels C915 and C917 must be placed in BYPASS B. (1) High RPV Level only (2) Jumpers must be installed on Panels C915 and C917 C. (1) High Main Steam Tunnel Temperature only (2) Jumpers must be installed on Panels C915 and C917 PNPS 2014 NRC Written 022014.doc
D. (1) High RPV Level and high Main Steam Tunnel Temperature (2) The PCIS RX WTR LVL Bypass switches on Panels C915 and C917 must be placed in BYPASS AND Jumpers must be installed on Panels C915 and C917 Proposed Answer: B Explanation (Optional):
A. Incorrect: Jumpers must be installed to defeat this isolation. The PCIS Bypass Switches only defeat the Low RPV level isolation (-46 inches). Therefore jumpers must be installed lAW Attachment 2 of PNPS 5.3.21. Plausible due to the name of the switches and the candidate believes that both water level isolation signals are bypassed by these switches.
B. Correct: An RPV High Water level isolation signal is present due to RPV level being >
+55 inches, Mode Switch out of Run with a low pressure condition. lAW Attachment 2 of PNPS 5.3.21, jumpers must be installed to defeat this isolation.
C. Incorrect: A Main Steam Tunnel Isolation does not exist and a high level isolation does exist. Plausible in that the steam tunnel temperature is above the isolation setpoint for the Turbine basement Exhaust (155 oF) but not the Steam Tunnel setpoint of 175 °F.
Additionally, it also plausible that the candidate may not recognize the high level isolation as this isolation signal is automatically bypassed under specified conditions.
D. Incorrect: A Main Steam Tunnel Isolation does not exist and a high level isolation does exist. Plausible in that the steam tunnel temperature is above the isolation setpoint for the Turbine basement Exhaust (155 oF) but not the Steam Tunnel setpoint of 175 °F.
The required action is plausible if the candidate believes that a high temperature isolation exists and that the PCIS Water level bypass switches bypass the high level also.
Technical Reference(s): PNPS 2.2.125, Attachment 3 (Attach if not previously provided)
PNPS 5.3.21, Attachments 1 and 2.
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-08-1 0, EO 3, and 5 (As available)
Question Source: Bank #
Modified Bank#
PNPS 2014 NRC Written 022014.doc
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 262001 A2.04 Importance Rating 3.8 Ability to (a) predict the impacts of the following on the A. C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Types of loads that, if de-energized, would degrade or hinder plant operation.
Proposed Question: RO Question # 16 The plant is at rated conditions when the preferred Main Transformer Cooling power supply from B-4 is lost.
( 1) Which one of the following is correct regarding the Backup/Emergency power supply to the Main Transformer Coolers, AND (2) What action would be required if this the Backup/Emergency power supply was unavailable?
A. ( 1) The power supply to the coolers will automatically shift to the Backup/Emergency source from B-3.
(2) Reactor power would have to be reduced below the Main Transformer's self-cooled rating.
B. ( 1) The power supply to the coolers will automatically shift to the Backup/Emergency source from B-3.
(2) The Main Transformer would have to be removed from service.
C. (1) The power supply to the coolers must be manually shifted to the Backup/Emergency source from B-3.
(2) Reactor power would have to be reduced below the Main Transformer's self-cooled rating.
D. ( 1) The power supply to the coolers must be manually shifted to the Backup/Emergency source from B-3.
(2) The Main Transformer would have to be removed from service.
PNPS 2014 NRC Written 022014.doc
Proposed Answer: B Explanation (Optional):
A. Incorrect: The transformer would have to be removed from service as there is no self-cooled rating for the Main. Plausible in that both the Startup and Unit Aux transformers have self-cooled ratings.
B. Correct: The transformer cooling fans and oil pumps are fed from either of two station 480 VAC load centers. The normal source is from bus B-4 and the alternate/emergency source is from bus B-3. An automatic transfer scheme is provided to re-energize the coolers from the alternate/emergency feeder should the normal source voltage fail.
Since the transformer does not have a self-Goaled rating, it must be removed from service.
C. Incorrect: The power supply shifts automatically to the backup power supply.
Additionally, the transformer must be removed from service.
D. Incorrect: The power supply shifts automatically to the backup power supply.
Technical Reference(s): Main Generator reference text, (Attach if not previously provided) pages 18 and 19 PNPS 2.2.2 page 11 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-01-07, E0-5 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 400000 A3.01 Importance Rating 3.0 Ability to monitor automatic operations of the CCWS including: Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS Proposed Question: RO Question # 17 A small steam leak in the drywall results in drywall pressure rising to 3.0 psig.
The reactor scrams but the Startup Transformer locks out when house loads attempt to transfer. Current electric plant conditions are as follows:
- The "A" EDG failed to start
- 4160 VAC bus A5 energized by the Shutdown Transformer
- 4160 VAC bus A6 energized by the "B" EDG Which one the following describes impact of the above on the TBCCW system?
A. Neither TBCCW pump is available due to the load shed signals.
B. The "A" TBCCW pump will auto start on low pressure 20 seconds after TBCCW pressure was lost.
The "B" TBCCW pump is not available due to the load shed signal.
C. The "B" TBCCW pump will be inhibited from starting until all ECCS pumps are running and then will restart on low system pressure.
The "A" TBCCW pump is not available due to the load shed signal.
D. The "A" TBCCW pump will be inhibited from starting until all ECCS pumps are running and then will restart on low system.
The "B" TBCCW pump is available for auto start if the "A" TBCCW pump fails.
Proposed Answer: A Explanation (Optional):
A. Correct: ER02116887 modified the load shed logic for TBCCW Pumps A and B (P-110A and P-1108). This modification changed the load classification of TBCCW Pumps A and B (P-110A and P-1108) from a Group 3 to a Group 21oad shed signal. Following the loss of 345kV offsite AC power or a degraded voltage on the Startup Transformer, coincident with a LOCA initiation signal, P-11 OA and P-11 08 will not auto-start. Diesel PNPS 2014 NRC Written 022014.doc
Load Shedding was initiated on A6 when the "B" EDG powered the bus in conjunction with the high drywell pressure condition. Diesel Load Shedding was initiated on A5 when due to the SUT supply breaker being open, the UAT supply open and a high drywell pressure signal.
B. Incorrect: Neither pump is available. Plausible if the candidate does not realize that the SOT powering the bus would also result in an EDG load shed signal.
C. Incorrect: Neither pump is available. Plausible if the candidate does thinks the TBCCW pumps are Group 3 loads (as they were in the past) and restart following a time delay to allow the LP ECCS pumps to start.
D. Incorrect: Neither pump is available. Plausible if the candidate does not understand that a load shed condition is present on both buses.
Technical Reference(s): 2.4.41, Discussion item #5 (Attach if not previously provided)
Figure 12 of Emergency AC reference Text.
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank#
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 261000 A3.03 Importance Rating 3.0 Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Valve operation Proposed Question: RO Question # 18 With the "A" Standby Gas Treatment train in its normal standby lineup, an operator places the control switch for the "A" SBGT fan to the RUN position.
Which one of the following is correct?
The "A" SBGT fan will ...
A. start and its outlet and inlet dampers will open.
B. not start until the operator manually opens the outlet and inlet dampers.
C. start, the outlet damper will open, and a low flow alarm will result due to the inlet damper not being opened by the operator.
D. not start until the operator establishes a complete suction flow path to the fan from the contaminated exhaust duct or from the drywell or torus.
Proposed Answer: A Explanation (Optional):
A. Correct: The fan will start. After the fan starts, the inlet and outlet dampers will open.
B. Incorrect: The fan will start. Plausible in that the "B" train is started by first opening the inlet damper which will in turn start the "B" fan. The "B" fan starting will then open the outlet damper.
C. Incorrect: The inlet damper will open. Plausible if the candidate does not understand that the fan starting will open the inlet damper.
D. Incorrect: The fan will start as described in "A" and both the inlet and outlet dampers will open. There is no interlock preventing system start if a complete suction path is not first started.
PNPS 2014 NRC Written 022014.doc
Technical Reference(s): SBGT Reference Text ,page 16 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-08-03, EO 10 (As available)
Question Source: Bank # PNPS 3929 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 212000 A4.05 Importance Rating 4.3 Ability to manually operate and/or monitor in the control room: Reactor power (RPS)
Proposed Question: RO Question # 19 The plant is at 100% power with a core flow of 69 Mlbm/hr when the following indications are received:
- Alarm Recirc Flow Converter Failure, C904R-A 1, annunciates
- The output of Recirc Flow Converter "A" is observed to be zero Assuming no manual action is taken, which one of the following is correct?
APRM "A" channel Hi-Hi setpoints have_ (1) _ _ (increased, decreased, remained the same).
The result of this event will be a _ ( 2 ) _ .
A. ( 1) decreased (2) Y2 scram and rod block B. (1) decreased (2) rod block only C. ( 1) increased (2) Y2 scram and rod block D. (1) remained the same (2) rod block only Proposed Answer: A Explanation (Optional):
A. Correct: The "A" flow Converter has failed downscale. This results in a zero core flow input to aii"A" side APRMs. This will cause the APRM scram and rod block setpoints to lower to those associated with a zero core flow. The scram setpoint for this condition is
- 70% with the rod block setpoint at- 63%. With power at 100%, a Y2 scram will occur.
PNPS 2014 NRC Written 022014.doc
A rod block will occur due to both an APRM Hi and a Flow Converter Downscale trip.
B. Incorrect: A Y:z scram will be generated. Plausible in that only a rod block would be generated if the flow converter had failed upscale.
C. Incorrect: Scram setpoints would decrease. Plausible if the candidate is confused as to the response of the flow biased settings on the APRMs as this would be true if the flow converter had failed upscale. Additionally the candidate may think that Recirc flow converter failure may generate a Y:z scram due to the flow converter going lnop.
D. Incorrect: Scram setpoints would decrease. Plausible if the candidate focuses on the high core flow and the APRM scram clamp at 118% for any core flow above 55Mibm/hr which prevents any increase in flow biased scram setpoints above 55Mibm/hr. If so then only a rod block would occur due to the flow converter failure.
Technical Reference(s): Pilgrim Power to Flow Map (Attach if not previously provided)
Proposed References to be provided to applicants during examination: Power to Flow Map Learning Objective: 0-R0-02-07-04, EO 5d, e (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 218000 A4.07 Importance Rating 3.5 Ability to manually operate and/or monitor in the control room: ADS valve acoustical monitor noise: Plant-Specific Proposed Question: RO Question # 20 EOP-16, RPV Flooding is in progress.
- All SRVs have been opened
- Flooding is in progress using the condensate system Which one of the following would indicate that the RPV has been flooded to the level of the Main Steam lines?
A. RPV pressure lowers to 30 psig and stabilizes.
B. Torus water level rises initially and then stabilizes.
C. SRV tailpipe temperatures reach saturation temperature for the given RPV pressure.
D. SRV acoustic monitor indications extinguish as RPV pressure lowers and then sporadically re-illuminate.
Proposed Answer: D Explanation (Optional):
A. Incorrect: This is an indication that the SRVs have closed. They are designed to stay physically open down to -50 psig. A stable pressure would indicate that they have not yet been forced open by raising pressure and forcing them to re-open.
B. Incorrect: Since a system external to the containment is being used, torus water level would continue to rise. Plausible in that this would be an indication if only systems aligned to the torus were being used.
C. Incorrect: When the main steam lines are flooded and the SRVs are now passing water, the tail pipe temperatures would lower to subcooled values.
D. Correct: As pressure lowers, the acoustic monitor lights would begin to extinguish as they are passing less flow. When the SRVs close at- 30 to 50 psig, they would PNPS 2014 NRC Written 022014.doc
extinguish. Once level was raised and the pressure began to rise they would be forced back open. However PNPS 5.3.35 cautions that the acoustic monitors may respond erratically to subcooled flow.
Technical Reference(s): PNPS 5.3.35, section 4.3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank#
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 262002 2.1.7 Importance Rating 2. 7 Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (UPS)
Proposed Question: RO Question# 21 Given the following:
- A sustained loss of 120 VAC Vital Bus Y-2 occurred
- A manual reactor scram is inserted.
Following the scram the following indications are observed:
- The C905 Group Scram Logic Lights de-energized
- 135 Blue scram lights illuminated
- The following alarms have annunciated:
o SPVAH Pressure Lo, C905R-F1 o RPIS INOP C905L-D4, o RWM ROD BLOCK" C905L-A4
- After several minutes all IRMs indicate -50 on Range 10 using back panel indications Which of the following methods would be effective in inserting any withdrawn control rods?
Method 1: Venting the scram air header.
Method 2: Venting the over-piston area of control rods.
Method 3: Inserting control rods by using the Continuous Insert feature of the Rod Movement Control Switch Method 4: Inserting control rods by using the Emergency In Feature of the Emergency In I Notch Override Switch A. Method 2 only B. Methods 1 and 3 only PNPS 2014 NRC Written 022014.doc
C. Methods 1 and 4 only D. Methods 2 and 4 only Proposed Answer: A Explanation (Optional):
A. Correct - The only method that would insert a control rod for these indications is venting the over-piston area of the control rods. With the Group Scram Logic Lights de-energized and the Blue scram lights illuminated this indicates a hydraulic failure has occurred. Venting the air header would be ineffective as it is already vented. Methods 3 and 4 would be ineffective as with a loss of the vital bus Y2 there is no rod select power available.
B. Incorrect: Method 1 would be ineffective as the air header is already vented.
Method 3 is not available as with Y-2 de-energized there is no rod select power.
C. Incorrect: Method 1 would be ineffective as the air header is already vented.
Method 4 is not available as with Y-2 de-energized there is no rod select power.
D. Incorrect: With a loss of the vital bus Y2 there is no rod select power available.
Technical Reference(s): 5.3.23, Sect 3.1, pg 6 (Attach if not previously provided) 5.3.6, Sect. 5.0 [9] pg 6 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-04, E0-23 (As available)
Question Source: Bank # PNPS 3191 Modified Bank #
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 PNPS 2014 NRC Written 022014.doc
55.43 Design, components, and function of reactivity control mechanisms and instrumentation.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 400000 2.1.2 Importance Rating 4.1 Conduct of Operations: Knowledge of operator responsibilities during all modes of plant operation. (Component Cooling Water)
Proposed Question: RO Question # 22 A major winter storm has hit the plant site resulting in the total loss of intake structure equipment.
The plant has been scrammed and offsite power is still available. Plant parameters are stable and in their normal band following the scram.
Which one of the following is an approved method of maintaining drywell cooling that can be implemented after the scram lAW PNPS 5.3.3 "Loss of all Service Water"?
A. Place RHR Loop 'A' in torus cooling and cross tie RBCCW Loops, "A" RBCCW Loop supplying.
B. Place RHR Loop 'A' and RHR Loop "B' in torus cooling. RBCCW Loop 'A' will cool both loops of RHR.
C. Establish public water supply to the fire water connection on the on the RBCCW supply to the RBCCW Loop "B" heat exchanger.
D. Place RHR Loop 'A' in torus cooling and establish a Fire Water connection directly to the RBCCW side of the RHR Loop 'A' heat exchanger.
Proposed Answer: A Explanation (Optional):
A. Correct. Per PNPS 5.3.3 "Loss of all Service Water" this method has been approved as an immediate way to maintain drywell cooling using the Torus as heat sink for drywell heat loads B. Incorrect. Only one loop of RHR is placed in service and both loops of RBCCW are cross-tied.
PNPS 2014 NRC Written 022014.doc
C. Incorrect. Fire water connection is on the SSW side of the heat exchanger.
D. Incorrect. RHR is not needed for this lineup and the fire water connection is on the SSW side of the heat exchanger.
Technical Reference(s): PNPS 5.3.3, Section 5.2 [1] (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-02-02, EO 9 (As available)
Question Source: Bank # X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215004 A2.05 Importance Rating 3.3 Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM)
SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/system Proposed Question: RO Question # 23 A Reactor Startup is in progress with the following conditions:
- Reactor is critical on a stable, positive period.
- Reactor power is on range 3 of the IRMs.
- SRM detector "A" is stuck and cannot be withdrawn.
- SRM "A" indication rises to 5E5 cps.
Given the above, which one of the following is correct regarding,
( 1) Any automatic actions AND (2) Any required actions directed by plant procedures that would allow the startup to continue?
A. ( 1) SRM rod block and upscale alarm.
(2) Bypass SRM "A" B. (1) SRM rod block and upscale alarm.
(2) Place SRM "A" function switch to standby.
C. (1) Upscale alarm only.
(2) Bypass SRM "A".
D. ( 1) Upscale alarm only.
(2) Place SRM "A" function switch to standby.
Proposed Answer: A PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Correct: If any SRM channel detects > 1 X 105 cps, a rod block is activated. This rod block is bypassed if:
- The mode switch is in RUN
-The associated IRM range switches are at range 8 or above Since the IRMs are on range 3, the rod block would occur.
lAW the ARP, the directed action is to bypass the SRM.
B. Incorrect: Taking the function switch out of operate will cause an INOP trip and associated rod block. Plausible if the candidate does not remember that taking the function switch out of Operate will cause an INOP trip. Additionally this action is not directed by any plant procedure.
C. Incorrect: A rod block also occurs. This answer would be correct if IRMs were on range 8 or higher. The candidate that confuses the range 8 IRM interlock with the range 3 interlock and does not understand the effect of the function switch may choose this answer. Additionally this action is not authorized by the procedure.
D. Incorrect: A rod block would also occur. Additionally this action is not authorized by the procedure.
Technical Reference(s): PNPS 2.2.64, Section 4.3 (Attach if not previously provided)
ARP C905L - E9 Proposed References to be provided to applicants during examination: None Learning Objective: SOURCE RANGE MONITORING, EO (As available) 6 Question Source: Bank # WTSI4904 Modified Bank #
New Question History: Last NRC Exam: Cooper 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PNPS 2014 NRC Written 022014.doc
10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 262001 A3.04 Importance Rating 3.4 Ability to monitor automatic operations of the A. C. ELECTRICAL DISTRIBUTION including:
Load sequencing Proposed Question: RO Question # 24 A loss of offsite power has occurred coincident with a loss of coolant accident.
The "B" diesel generator has started and has re-energized 4160 VAC Bus A-6.
Which of the following represents the correct order for the sequencing of loads onto Bus A-6?
A. RHR Pump "B" starts RHR Pump "D" starts Core Spray Pump "B" starts SSW Pump "D" starts RBCCW Pump "D" starts B. Core Spray Pump "B" starts RHR Pump "B" starts RHR Pump "D" starts SSW Pump "D" starts RBCCW Pump "D" starts C. RHR Pump "B" starts RHR Pump "D" starts Core Spray Pump "B" starts RBCCW Pump "D" starts SSW Pump "D" starts D. Core Spray Pump "B" starts RHR Pump "B" starts RHR Pump "D" starts RBCCW Pump "D" starts SSW Pump "D" starts Proposed Answer: B PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: The Core Spray pump starts first.
B. Correct: The following sequence of events would automatically occur if a LOCA occurred simultaneously with a loss of off-site power.
Time Zero:
Diesel generator breakers close.
Start of both Core Spray pumps 5 Seconds From Time Zero:
Start of first two RHR pumps (A and B).
Core spray pumps at speed.
10 Seconds From Time Zero:
Start of next two RHR pumps (C and D).
First two RHR pumps at speed.
20 Seconds From Time Zero:
SSW pump D starts(< 2.3 # w/a 20 second time delay).
Vital AC MG set returns to AC power.
25 Seconds From Time Zero SSW pump A starts if pressure is not normal (<2.3 # w/a 25 second time delay).
30 Seconds From Time Zero RBCCW pumps A and D start(< 60 # w/a/ 30 second time delay)
C. Incorrect: The Core Spray pump starts first. The SSW pump starts before the RBCCW pump D. Incorrect: The SSW pump starts before the RBCCW pump.
Technical Reference(s): PNPS 2.4.16, Attachment 7 (Attach if not previously provided)
Emergency AC Distribution reference text, pages 23 and 24 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-08 EO 11 (As available)
Question Source: Bank # WTSI12470 Modified Bank #
PNPS 2014 NRC Written 022014.doc
New Question History: Last NRC Exam: 2010 Riverbend Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215005 K5.06 Importance Rating 2.5 Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM : Assignment of LPRM's to specific APRM channels Proposed Question: RO Question # 25 The plant is at power. APRM "A" is currently reading 100%. There are no bypassed LPRMs.
Then, an LPRM associated with an APRM "A" fails downscale. No operator action has yet been taken.
Following the failure what will the APRM Back Panel meter indicate if the Meter Function switch is placed in the COUNT position?
A. 65 B. 70 C. 75 D. 80 Proposed Answer: D Explanation (Optional):
A. Incorrect: The meter will indicate 80 (see explanation for "D"). Plausible if the candidate confuses the number of LPRMs assigned to "A" side APRMs and "B" side APRMs and concludes that "A" APRM has only 14 LPRMs. Then if the candidate concludes that with the LPRM failing downscale there are only 13 remaining. However this would also be incorrect. Unlike the RBM which utilizes downscale trip units that will automatically remove a failed LPRM from its averaging circuit, this is not the case for the APRMs. If so this would result in an indication of 65 ( 13 X 5%=65).
B. Incorrect: The meter will indicate 80. Plausible if the candidate confuses the number of LPRMs assigned to "A" side APRMs and "B" side APRMs and concludes that "A" APRM has only 14 LPRMs. Then if the candidate correctly concludes that the COUNT circuit counts all LPRMs that have not been bypassed will conclude that the meter will read 70% (14 X5%=70%)
PNPS 2014 NRC Written 022014.doc
C. Incorrect: The meter will indicate 80. Plausible if the candidate correctly identifies that there are 16 LPRMs assigned to "A" APRM but assumes that the LPRM failing downscale will automatically be removed from the COUNT Circuit and that only 15 remain (15 X 5% =75)
D. Correct: There are 16 LPRMs assigned to APRM "A". The COUNT circuit determines the number of assigned LPRMs which are in OPERATE (5 percent per LPRM). Since the LPRM has not yet been bypassed, the indication will remain at 80 (16 X 5% =80).
Technical Reference(s): APRM Reference text Page 7 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New New Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215003 2.4.31 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of annunciator alarms, indications, or response procedures. (IRM)
Proposed Question: RO Question # 26 A reactor startup is in progress. All IRMs are at 50/125 on range 4.
Then, the following alarms are observed:
- Rod Withdrawal Block alarm, C905L-B4
- Neutron Monitoring Trip alarm, C905R-C3 Which one of the following inappropriate actions is consistent with these indications?
IRME ...
A. being down ranged to range 3.
B. being up ranged to range 5.
C. being withdrawn from the core.
D. function switch taken out of Operate.
Proposed Answer: A Explanation (Optional):
A. Correct: Since the odd numbered ranges are just an expanded view of the even number ranges, the IRM would try to indicate 50 on range 3. Range 3 has a scale of 0-40. This would result in exceeding both the Hi and Hi-Hi trip setpoints which produce the indicated alarms B. Incorrect: This would result in downscale indications as the IRM would indicate 5/125 of scale on range 5.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: This would result in downscale indications.
D. Incorrect: This would generate an INOP trip. An INOP trip would not result in the IRM Hi alarm.
Technical Reference(s): IRM Reference Text, pages 10 (Attach if not previously provided) and 15 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-07 -02, EO 14 (As available)
Question Source: Bank # WTSI1262 Modified Bank #
New Question History: Last NRC Exam: 2007 Fermi Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 204000 K1.14 Importance Rating 2.5 Knowledge of the physical connections and/or cause- effect relationships between REACTOR WATER CLEANUP SYSTEM and the following: Process sample system Proposed Question: RO Question # 27 The reactor is at rated conditions when the following occur:
- Steam Leakage Area Temp Hi, C904L-A6, alarms
- RWCU isolates
- The alarm is determined to be due to RWCU Heat Exchanger Room high area temperature What action is now required regarding Recirc Loop Sample Valves A0-220-44 and A0-220-45 and why?
The valves should be .....
A. Opened or verified open to aid in depressurizing the RWCU system.
B. Closed or verified closed to prevent damage to the Mitigation Monitoring System.
C. Closed or verified closed to prevent leakage from the reactor to the reactor building.
D. Opened or verified open to re-establish a flow of water to the RWCU conductivity elements.
Proposed Answer: C Explanation (Optional):
A. Incorrect: The valves are closed. Plausible since the indications provided are indicative of a RWCU leak. If the candidate remembers that there is an interface between the RWCU and the Mitigation Monitoring System (MMS) through the Recirc Loop Sample Valves the candidate may conclude that this is an appropriate action.
B. Incorrect: Plausible if the candidate remembers that there is an interface between the RWCU and the Mitigation Monitoring System through the Recirc Loop Sample Valves the candidate may conclude that this is an appropriate action.
PNPS 2014 NRC Written 022014.doc
C. Correct: Both the ARP for C904L-A6 and 2.4.47 direct that the Recirc Loop Sample valves be closed if RWCU isolates on high area temperature or flow. Isolation of the MMS is necessary if a valid steam leak detection signal has isolated the RWCU System. This is done by closing the Recirc Loop Sample valves. Valid steam leak detection signals are steam leakage area temp HI and/or inlet flow HI. This action prevents the MMS from providing a leakage pathway from the Reactor to the Secondary Containment.
D. Incorrect: The valves are closed.
Technical Reference(s): ARP C904L-A6 (Attach if not previously provided)
PNPS 2.4.27, Subsequent Action
[1] and discussion item [3]
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-05, 20a (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 226001 K1.08 Importance Rating 3.2 Knowledge of the physical connections and/or cause-effect relationships between RHRILPCI:
CONTAINMENT SPRAY SYSTEM MODE and the following: Nuclear boiler instrumentation Proposed Question: RO Question # 28 A small break LOCA in the drywell has occurred.
Current plant conditions are as follows:
- 'A' Loop of RHR is has been placed in drywell spray mode lAW PNPS 5.3.35.2, Attachment 16, Drywell Spray Checklist
- Drywell pressure is currently 6 psig and slowly lowering.
- Reactor pressure is 420 psig and slowly lowering.
- RPV level is -40 inches and slowly lowering.
Without further operator action, RHR pump flow will remain aligned for drywell spray until:
A. Reactor level lowers to -46 inches B. Reactor level lowers to -150 inches C. Drywell pressure lowers to 2.2 psig D. Reactor pressure lowers to 400 psig Proposed Answer: B Explanation (Optional):
A. Incorrect: In order to establish drywell sprays initially the LPCI initiation signal had to be placed in override because of the high drywall pressure condition. RPV level lowering to -46 inches will have no additional impact.
B. Correct: Although the LPCI initiation signal has been overridden, when RPV level lowers to the 2/3rds coverage interlock, -150 inches, the valves will close until the 2/3rd core coverage interlock is overridden.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: Although a lowering drywell pressure will eventually close the spray valves, this will not occur until drywell pressure lowers to 1.8 psig. Plausible in that 2.2 psig is the LPCI initiation setpoint.
D. Incorrect: The drywell spray valves will remain open when pressure lowers to 400 psig.
Plausible in that the LPCI system will realign for LPCI injection at this pressure.
However since the LPCI initiation signal was already overridden the valves will remain open.
Technical Reference(s): RHR Reference Text, pages 44- (Attach if not previously provided) 46.
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-01, EO 14 (As available)
Question Source: Bank # WTSI #2172 Modified Bank #
New Question History: Last NRC Exam: Pilgrim 2002 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 215002 K3.01 Importance Rating 3.3 Knowledge of the effect that a loss or malfunction of the ROD BLOCK MONITOR SYSTEM will have on following: Reactor manual control system: BWR-3,4,5 Proposed Question: RO Question # 29 The plant is at 90% power with a center control rod selected.
Then, the averaging circuit for RBM Channel "A" malfunctions such that the output of the RBM lowers to 84%.
Which one of the following is correct?
A. No RBM trips will occur. No alarms or rod blocks will occur.
B. A RBM INOP trip will occur generating a rod withdraw block and associated alarms.
C. A RBM DOWNSCALE trip will occur generating a rod withdraw block and associated alarms.
D. A RBM DOWNSCALE trip and associated alarm will occur. A rod withdraw block will not occur.
Proposed Answer: C Explanation (Optional):
A. Incorrect: A DOWNSCALE trip will occur. Plausible if the candidate realizes that the condition will not result in an INOP trip and confuses the downscale trip setting with that of all other instruments of - 5%.
B. Incorrect: A DOWNSCALE trip will occur. Plausible if the candidate believes that an INOP trip will occur for this condition.
C. Correct: A RBM downscale trip occurs at 94% and will result in a rod block, a downscale alarm and rod block alarm.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: A rod block will occur.
Technical Reference(s): RBM Reference text, Figure 2. (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-07-05, EO 2h (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 201002 K4.07 Importance Rating 2.5 Knowledge of REACTOR MANUAL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Timing of rod insert and withdrawal cycles (rod movement sequence timer)
Proposed Question: RO Question # 30 During a single notch withdraw of a control rod, the RMCS timer applies a withdraw signal for 3.0 seconds.
Which one of the following is correct regarding future rod movement?
A. The rod will settle into the next notch. Additional withdrawals are permissible provided they are within the withdraw limits of the RWM.
B. A rod block will terminate the rod withdraw. Withdrawal of any other rod is also blocked.
Insertion of any rod is allowed.
C. The rod will be de-selected preventing further withdrawal. All other rods can be inserted or withdrawn as required.
D. The rod will be de-selected preventing further withdraw. Rod selection of all remaining rods is blocked.
Proposed Answer: D Explanation (Optional):
A. Incorrect: A select block will be generated which will de-select the rod. Plausible if the candidate believes that 3 seconds is within the limits of the Aux timer.
B. Incorrect: A select block is generated. Plausible if the candidate knows withdraw will be prevented but does not understand that a select block is generated.
C. Incorrect: All other rods are prevented from being inserted or withdrawn due to the inability to select any rod.
D. Correct: The normal notch out withdraw portion of the sequence takes 1.5 seconds. If a withdraw signal is sent to directional control valves for more than 2 seconds, an PNPS 2014 NRC Written 022014.doc
auxiliary timer times out. When the auxiliary timer (in panel 928) times out, it generates a select block. This deselects the selected rod and prevents further rod selection until the block has been cleared. This prevents a faulty master timer from causing an uncontrolled continuous withdrawal signal.
Technical Reference(s): RMCS Reference Text, page 15 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-08, EO 1Oa (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 201001 K5.02 Importance Rating 2.6 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM : Flow indication Proposed Question: RO Question# 31 The plant is at rated conditions when the operator notes that the CRD flow indication on the CRD Flow controller is 75 gpm.
Which one of the following is a potential adverse consequence of this indication?
A. Control rods drifting in.
B. Control rods drifting out.
C. Control rod accumulator Trouble Alarms.
D. Control rod drive filter break through and clogging of the directional control valves.
Proposed Answer: A Explanation (Optional):
A. Correct: PNPS 2.4.11, Attachment 2 directs that if a rod is drifting in then check that the cooling water differential pressure is not too high. A CRD Flow pegged high would result in excessive cooling water differential pressure. Since cooling water flow is felt on the underside of the CRDM drive piston this could result in a rod drifting into the core.
B. Incorrect: Rods will not drift out. The cooling water pressure is felt on the underside of the drive piston.
C. Incorrect: An accumulator will alarm if pressure drops to 965 psig. Although pressure will drop in the CRD system it will not drop any lower than reactor pressure - otherwise flow would not increase. Reactor pressure at rated conditions is significantly higher than 965 psig.
D. Incorrect. A high dp alarm would be received prior to this occurring. Additionally, there is a Y- Strainer downstream of the filter to prevent clogging of the directional control valves.
PNPS 2014 NRC Written 022014.doc
Technical Reference(s): PNPS 2.4.11, Attachment 2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-11, EO 2g (As available)
Question Source: Bank #
Modified Bank# PNPS Bank 2828 Modified stem so that FCV fails open not closed.
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 214000 K6.02 Importance Rating 2.7 Knowledge of the effect that a loss or malfunction of the following will have on the ROD POSITION INFORMATION SYSTEM : Position indication probe Proposed Question: RO Question # 32 While withdrawing a control rod from position 24 to 26, reed switch 25 fails to open.
The rod is presently latched at notch 26. The expected indication(s) for this condition is/are:
A. Black-black (no indication) for the rod on the four-rod display ONLY.
B. Red drift light lit on the full core display, "25" displayed on the four-rod display.
C. Red drift light lit on the full core display and black-black (no indication) for the rod on the four-rod display.
D. Red drift light lit on the full core display, both "25" and "26" superimposed on one another on the four-rod display.
Proposed Answer: D Explanation (Optional):
A. Incorrect: This would be the indication if the position 26 reed switch failed to close.
Plausible if the candidate does not understand whether reed switches are normally open or closed.
B. Incorrect: "25" and "26" would be super imposed over one another.
C. Incorrect: "25" and "26" would be super imposed over one another on the four-rod display.
D. Correct: The drift circuit will annunciate if a control rod is NOT selected and driving and either its even numbered reed switch opens, or an odd numbered reed switch closes.
Since the position "25" reed switch is stuck closed, a drift alarm will occur after the RMCS timer times out. Since two switches are closed, the two numbers will be superimposed over one another.
PNPS 2014 NRC Written 022014.doc
Technical Reference(s): RPIS Reference Text, page 8 and (Attach if not previously provided) 13 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-04, E0-7 (As available)
Question Source: Bank # 4927 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 202001 A1.06 Importance Rating 2.5 Ability to predict and/or monitor changes in parameters associated with operating the RECIRCULATION SYSTEM controls including: Recirculation pump motor amps Proposed Question: RO Question # 33 Given the following:
- The plant is at rated power
- 125 VDC bus D-5 is lost
- Two minutes later the reactor operator reports that current on the "B" Recirc MG set motor is pegged high and that P225B is not running Which one of the following actions is required by PNPS 5.3.12, LOSS OF ESSENTIAL DC BUS D17 OR D5 AND D37?
A. Manually scram the reactor.
B. Manually trip 4KV bus A4 at panel C3.
C. Locally trip the "B" Recirc MG Drive motor breaker at 4KV bus A4.
D. Locally start the "B" Recirc System DC Emergency Bearing Oil Pump at 250 VDC panel D9.
Proposed Answer: A Explanation (Optional):
A. Correct- Per the procedure, if there is indication of severe bearing damage (locked rotor condition), then a Scram is initiated and the resulting Turbine trip will de-energize the Unit Auxiliary Transformer, which in turn results in de-energizing Bus A4 and thereby securing the "B" Recirc MG Set.
PNPS 2014 NRC Written 022014.doc
B. Incorrect: P225B tripped as expected following the loss of D5. With no AC lube oil pumps running the MG set is seizing. However with the loss of D5, bus A4 is also without control power.
C. Incorrect- The procedure directs that the MG set be tripped locally but only if the current is NOT pegged high due to the electrical hazard of tripping the breaker under such a high load.
D. Incorrect: Starting the DC pump will not alleviate the condition as it only supplies oil to the coupler bearings and not the motor and generator bearings.
Technical Reference(s): PNPS 5.3.12 LOSS OF (Attach if not previously provided)
ESSENTIAL DC BUS D17 OR D5 AND D37, discussion section, item
[3]
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: 2011 Pilgrim # 56 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 286000 A2.12 ImportancE~ Rating 3.1 Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low diesel fuel supply:
Proposed Question: RO Question # 34 The plant is experiencing station blackout conditions. The Diesel Driven Fire Pump is being used for vessel makeup.
Assuming that the Diesel Driven Fire Pump is being run at full capacity, (1) How long will the diesel continue to run without any operator action before it shuts down due to lack of fuel?
AND (2) What action can be taken to extend the time available?
A. (1) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (2) Manually align the diesel oil transfer pump (P-141 A) and transfer diesel fuel from the EDG fuel oil tanks to the diesel fire pump.
B. (1) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (2) Manually align the hydro-turbine fuel oil transfer pump (P-181) and transfer diesel fuel from the EDG fuel oil tanks to the diesel fire pump.
C. (1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) Manually align the diesel oil transfer pump (P-141A) and transfer diesel fuel from the EDG fuel oil tanks to the diesel fire pump.
D. (1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) Manually align the hydro-turbine fuel oil transfer pump (P-181) and transfer diesel fuel from the EDG fuel oil tanks to the diesel fire pump.
Proposed Answer: B Explanation (Optional):
PNPS 2014 NRC Written 022014.doc
A. Incorrect: The hydro-turbine fuel oil transfer pump (P-181) must be used. Plausible in that P-141 A is normally used to transfer fuel oil from the EDG storage tanks to the Diesel Fire Pump day tank. However this pump is not DC powered and would not be available during blackout conditions.
B. Correct: The diesel fire pump day tank is manually kept full from the EDG oil transfer system. A full day tank provides continuous operation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During loss of all AC, the fuel oil transfer system will be configured to operate the hydro-turbine fuel oil transfer pump per PNPS 2.4.54. The hydro-turbine pump does not require any power to operate. It uses the discharge from the fire pump to run a turbine which operates the transfer pump. The hydro-turbine fuel oil transfer pump is not for fire protection, but for "beyond design basis accidents" when the fire pump is used for Rx vessel makeup.
C. Incorrect: A full day tank provides continuous operation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Plausible in that it is a "day" tank. Also The hydro-turbine fuel oil transfer pump (P-181) must be used.
D. Incorrect: A full day tank provides continuous operation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. However this pump is not DC powered and would not be available during blackout conditions.
Technical Reference(s): Fire Water Protection System Ref (Attach if not previously provided) text pages 15, 16 and 39 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-10-05, 5d, 9 (As available)
Question Source: Bank#
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 288000 A3.01 Importance Rating 3.8 Ability to monitor automatic operations of the PLANT VENTILATION SYSTEMS including:
Isolation/initiation signals Proposed Question: RO Question # 35 The plant is refueling. An air purge of the drywell through the Contaminated Exhaust Fans is in progress.
Then, the following process rad monitor Upscale trips occur:
- Channel "A" Refuel Floor Exhaust RM 1705-*8A
- Channel "B" Refuel Floor Exhaust RM 1705-*8B Given that there are NO other tripped PRMs which one of the following is correct regarding any automatic actions that may or may not occur?
A. NO isolations occur B. Complete Reactor Building Ventilation isolation Drywell purge lineup does NOT isolate C. Complete Reactor Building Ventilation isolation Complete Drywell purge lineup isolation D. One of the two in-series Reactor Building Isolation Dampers isolate One of the two in-series Drywell Purge supply and exhaust valves isolate Proposed Answer: C Explanation (Optional):
A. Incorrect: Both the Reactor Building and the Purge lineup isolated. Plausible if the candidate does not understand the configuration of the PRMs in the RBIS logic and concludes that the RBIS logic does not actuate as there are some combinations of PRM trips that will not cause an isolation.
B. Incorrect: The drywell purge lineup isolates. Plausible if the candidate does not realize PNPS 2014 NRC Written 022014.doc
that an RBIS trip also isolates the Primary Containment Atmosphere Control Valves (PCAC).
C. Correct: Refuel Floor Exhaust Channels A & B tripping will result in both RBIS logic train actuations this will cause a Reactor Building Ventilation isolation and a PCAC isolation.
D. Incorrect: All Reactor Building and Drywell purge dampers will isolate.
Technical Reference(s): PCAC reference text page 24 & (Attach if not previously provided)
Figures 9 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-08-05, EO 10, 11 (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 241000 A4.11 Importance Rating 3.1
---=-----:-:=
Ability to manually operate and/or monitor in the control room: Turbine speed (Reactor/Turbine Pressure Regulator)
Proposed Question: RO Question # 36 The Main Turbine is currently on the jack and is about to be rolled during a plant startup.
The Turbine is rolled off the jack and the initial Turbine acceleration is controlled using the_ (1) _.
While the Turbine speed is increasing, Reactor Pressure is maintained by the
_(2)_.
A. (1) Load Limit Control Switch.
(2) Bypass Valves B. ( 1) Speed/Load Changer Control Switch (2) Turbine Control Valves C. (1) Load Limit Control Switch.
(2) Turbine Control Valves D. (1) Speed/Load Changer Control Switch (2) Bypass Valves Proposed Answer: A Explanation (Optional):
A. Correct: The turbine is initially rolled using the load limit which is initially set to zero. As the control valves open slightly to accelerate the turbine the bypass valves close down to control pressure.
B. Incorrect: The turbine is rolled with the Load Limit and pressure is controlled by the bypass valves. Both are plausible. The Speed load changer will limit the speed of the turbine to - 1710 RPM and is used to complete the acceleration of the turbine to rated speed. Since the control valves are opening and the control valves normally throttle to control pressure a candidate may assume that this is also true during the initial roll.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: The bypass valves will throttle as required to control pressure.
D. Incorrect: The load limit is used initially to roll the turbine.
Technical Reference(s): PNPS 2.1.1, steps [111] thru [115] (Attach if not previously provided) and [127]
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-05-04, 13e, f, 14a (As available)
Question Source: Bank# X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 219000 2.1.32 Importance Rating 3.8 Conduct of Operations: Ability to explain and apply all system limits and precautions (RHRILPCI: Torus/Pool Cooling Mode)
Proposed Question: RO Question # 37 The plant is at rated conditions.
"A" Loop of RHR is placed in Torus Cooling in preparation for HPCI Testing.
lAW 2.2.19, Residual heat Removal, which one of the following Tech Spec actions is required prior to placing "A" RHR into Torus Cooling and why?
A. Declare Primary Containment inoperable. If a loss of offsite power was to occur and the "A" EDG failed to start, the Torus Cooling valves will be inoperable in the open position.
B. Declare LPCI inoperable. If a loss of offsite power was to occur and the "A" EDG failed to start coincident with a LOCA, the Torus Cooling valves will lose power and remain open, diverting injection flow to the Torus.
C. Declare LPCI inoperable. If a loss of offsite power was to occur while aligned for Torus Cooling the RHR loop will drain and create a potential for subsequent water hammer if an RHR pump was started.
D. Declare "A" Loop of Torus Cooling Inoperable. If a loss of offsite power was to occur and the "A" EDG failed to start, the Torus Cooling valves will be inoperable in the open position.
Proposed Answer: B Explanation (Optional):
A. Incorrect: LPCI mode is declared inoperable. Plausible in that the situation described would be true.
B. Correct: lAW of PNPS 2.2.19, section 4.1.2 item [3], the requirement to enter an Active LCO for the LPCI System being inoperable (Tech Spec 3.5.A) whenever the LPCI System is in the Torus Cooling mode of operation is based upon a postulated accident scenario that involves a LOOP-LOCA while in Torus Cooling and a single failure of an PNPS 2014 NRC Written 022014.doc
EDG. With the loss of the EDG, the Torus Cooling valves will lose power and remain open, diverting flow to the Torus during the subsequent LPCI injection with the remaining two LPCI pumps.
C. Incorrect: Plausible in that this would occur but is not the reason for declaring the system inoperable because the system is designed so that significant damage would not occur due to water hammer. TS Bases, page B3/4.5-24 D. Incorrect: LPCI is declared inoperable.
Technical Reference(s): PNPS 2.2.19, section 4.1.2 item (Attach if not previously provided)
[3].
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-01, EO 26 (As available)
Question Source: Bank # WTS 11830 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: 2010 NMP2 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 259001 K4.04 Importance Rating 2.5 Knowledge of REACTOR FEEDWATER SYSTEM design feature(s) and/or interlocks which provide for the following: Dispersal of feedwater in the reactor vessel Proposed Question: RO Question # 38 The reactor is at rated conditions. The operator notes the following feedflow indications:
- "A" Feedwater flow, FI-640-24A indicates 4.25 Mlbm/hr
- "8" Feedwater flow, Fl-640-248 indicates 3. 75 Mlbm/hr How is the feedwater dispersed in the reactor vessel to ensure proper mixing of the feedwater and down comer flow?
AND Given the indications above how can this feedwater flow deviation be eliminated?
A. The two feed lines downstream of the feed reg valves split into 2 lines so there are 4 vessel penetrations equally spaced around the vessel. The feedwater flows directly into the annulus area mixing with the downcomer flow.
Equalize flow by adjusting the bias settings on the FWLC M/A Stations.
B. The two feed lines downstream of the feed reg valves split into 2 lines so there are 4 vessel penetrations equally spaced around the vessel. The feedwater flows directly into the annulus area mixing with the downcomer flow.
Equalize flow by placing one FWLC M/A Station in manual and adjusting as required.
C. The two feed lines downstream of the feed reg valves split into 2 lines so there are 4 vessel penetrations equally spaced around the vessel. The feedwater then flows through 4 spargers to distribute the flow in the annulus area.
Equalize flow by adjusting the bias settings on the FWLC M/A Stations.
D. The two feed lines downstream of the feed reg valves split into 2 lines so there are 4 vessel penetrations equally spaced around the vessel. The feedwater then flows through 4 spargers to distribute the flow in the annulus area.
Equalize flow by placing one FWLC M/A Station in manual and adjusting as required.
Proposed Answer: C PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: The feedflow is distributed to the downcomer annulus via feedwater spargers.
B. Incorrect: The feedflow is distributed to the downcomer annulus via feedwater spargers. Additionally, PNPS 2.2.82 directs using the bias settings to equalize flow.
C. Correct: Four 12-inch nozzles direct feedwater into the reactor vessel to replace the steam sent to the turbine. The feedwater nozzles discharge through the feedwater spargers to the downcomer annulus. PNPS 2.2.82 directs using the bias settings to equalize flow.
D. Incorrect: PNPS 2.2.82 directs using the bias settings to equalize flow.
Technical Reference(s): PNPS 2.2.82 Precaution 5.0 (Attach if not previously provided)
Condensate and Feed Reference Text page 30.
Reactor Vessel and Internals Reference Text, page 16 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-07, EO 5a (As available) 0-R0-02-04-10, EO 13 Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295023 AK1.01 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS : Radiation exposure hazards Proposed Question: RO Question# 39 Core Alterations are in progress when the following occurs:
- An irradiated fuel bundle is being moved from the reactor cavity to the Spent Fuel Pool
- The Bundle becomes ungrappled and falls into the reactor vessel downcomer area.
(Between the vessel wall and the shroud)
- Bundle integrity is maintained Which one of the following workers is at greatest risk of radiation overexposure?
A. I&C Tech at the SBLC tank B. Refuel SRO on the Bridge C. Mechanic working on SRVs D. RP Tech at Refuel Floor Access Point Proposed Answer: C Explanation (Optional):
A. Incorrect: because of the location of these components. SRO on the bridge is shielded by water level within the cavity, as is the RP Tech at the access. SLC Tank is in Secondary Containment, which is shielded by Primary Containment wall.
B. Incorrect: because of the location of these components. SRO on the bridge is shielded by water level within the cavity, as is the RP Tech at the access. SLC Tank is in Secondary Containment, which is shielded by Primary Containment wall.
C. Correct: because of the location of these components. SRO on the bridge is shielded by water level within the cavity, as is the RP Tech at the access. SLC Tank is in Secondary Containment, which is shielded by Primary Containment wall.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: because of the location of these components. SRO on the bridge is shielded by water level within the cavity, as is the RP Tech at the access. SLC Tank is in Secondary Containment, which is shielded by Primary Containment wall.
Technical Reference(s): Plant layout drawing (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank# WTS Bank 1048 Modified Bank#
New Question History: Last NRC Exam: 2008 Nine Mile Point 2 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295028 EK1.02 Importance Rating 2.9 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Equipment environmental qualification Proposed Question: RO Question # 40 The drywell has a maximum internal design temperature of_ (1) _and the ADS SRV solenoids are designed to operate up to an ambient drywell temperature of_ (2) _.
(1) Maximum Drvwell Internal (2) Maximum SRV Solenoid Temperature Temperature A 215°F 281°F B. 215°F 330°F C. 281°F 281°F D. 281°F 330°F Proposed Answer: D Explanation (Optional):
A Incorrect : 215°F is the limiting drywell temperature to guarantee that ECCS trips occur on/or before present T.S. values and requires a shutdown be initiated if drywell temperature cannot be restored to less than 215°F within 30 minutes. 281 oF is the maximum internal design temperature of the drywell.
B. Incorrect: 215°F is the limiting drywell temperature to guarantee that ECCS trips occur on/or before present T.S. values and requires a shutdown be initiated if drywell temperature is restored less than 215°F within 30 minutes.
C. Incorrect: The ADS SRV solenoids are designed to operate up to an ambient drywell temperature of 330°F.
D. Correct: lAW Primary Containment reference text Section C.1. , the maximum internal design temperature of the drywell is 281 degrees F. The ADS SRV solenoids are designed to operate up to an ambient drywell temperature of 330°F. Main Steam PNPS 2014 NRC Written 022014.doc
System reference text page 16 - Engineering determined that the ambient temperature inside the drywell during a main steam line break may rise to 330°F The design basis MSLB is the most limiting break. This would cause the nitrogen inside the accumulator to increase in pressure due to the lack of area for expansion. The pressure would continue to rise to 160 psi. The 160 psi pressure exceeds the design rating of the previous accumulator (135 psi) solenoid valves. For this reason, new solenoid valves and relief valves designed to operate with a maximum Nitrogen pressure of 160 psid were installed on the accumulators. The maximum internal design temperature of the drywell is 281 °F.
Technical Reference(s): Main Steam System reference text (Attach if not previously provided) page 14 Primary Containment Structure reference text page 8 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-04-01, E06 (As available) 0-R0-03-04-05, E011 Question Source: Bank# X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: Pilgrim 2011, # 44 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Shielding, isolation, and containment design features, including access limitations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295004 AK 1. 04 Importance Rating 2.8 AK1.04- Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Effect of battery discharge rate on capacity Proposed Question: RO Question # 41 Station blackout conditions have existed for the past 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Which of the following actions specified in PNPS 5.3.31, Station Blackout, will reduce the discharge rate on the "A" 125 VDC battery and extend the battery's capacity?
Action 1: Stopping Recirc MG set "A" DC Emergency Bearing Oil Pump Action 2: Stopping the Turbine Emergency Bearing Oil Pump Action 3: Transferring 125VDC bus D-6 to its alternate supply A. Action 1 only B. Action 3 only C. Actions 1 and 2 only D. Actions 2 and 3 only Proposed-Answer: B Explanation (Optional):
A. Incorrect: Action 1 is associated with a 250 VDC load. Plausible in that this is an action addressed in PNPS 5.3.31 for reducing overall loads on the DC distribution system.
B. Correct: lAW subsequent actions of PNPS 5.3.31, step [3] (e), if power has NOT been restored to MCC B15, THEN, AFTER approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, TRANSFER the 125V DC "swing bus" (D6) from the "A" 125V battery to the "B" 125V battery.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: Action 2 is associated with a 250 VDC load. Plausible in that this is an action addressed in PNPS 5.3.31 for reducing overall loads on the DC distribution system.
D. Incorrect: Actions 2 and 3 are associated with 250 VDC loads. Plausible in that these actions are addressed in PNPS 5.3.31 for reducing overall loads on the DC distribution system.
Technical Reference(s): PNPS 5.3.31, step [3] (e) (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 600000 AK2.01 Importance Rating 2.6 Knowledge of the interrelations between PLANT FIRE ON SITE and the following: Sensors, detectors and valves Proposed Question: RO Question # 42 The plant is at 100% when a Main Transformer Lockout results in a scram. A report of fire from the Main Transformer area is received.
Which one of the following systems should respond (1) AND if this system does not automatically respond what operator actions (2) are required?
A. ( 1) A Deluge Water System (2) Manually actuate from Panel C-7 B. (1) Preaction water spray system (2) Manually actuate locally from the Turbine Building Trucklock area C. ( 1) A Deluge water system (2) Manually actuate locally from the Turbine Building Trucklock area D. (1) Preaction water spray system (2) Manually actuated from Panel C-7 Proposed Answer: C Explanation (Optional):
A. Incorrect- Actuation is from the local Deluge station TB 23' Turbine Trucklock area B. Incorrect- it's a Deluge system that actuates following a Main Transformer lockout when "rate of rise" temperature indicators trip C. Correct - Deluge system actuates from following a Main Transformer lockout when "rate of rise" temperature indicators trip the Deluge valve. If the Deluge valve fails to trip, manual actuation can be accomplished locally at the Deluge station TB 23' Turbine Trucklock area PNPS 2014 NRC Written 022014.doc
D. Incorrect- Actuation is from the local Deluge station TB 23' Turbine Trucklock area Technical Reference(s): 5.5.2, Att. 18, pg 82, 85 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-10-05, Terminal Objective, (As available)
MANUALLY ACTIVATE FIRE PROTECTION SYSTEM.
Question Source: Bank# X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295019 AK2.09 Importance Rating 3.3 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Containment Proposed Question: RO Question # 43 The plant is at rated conditions with all systems aligned for normal full power operation.
Then, annunciator AIRIN2 TO DRYWELL TROUBLE, C904LC F-3, alarms.
Pneumatic supply pressure to the drywell on panel C7 is 102 psig and stable.
This means that Drywell pneumatics has automatically swapped to the ....
A. nitrogen storage tank B. backup nitrogen bottles C. instrument air system supply D. SRV backup nitrogen accumulator supply Proposed Answer: B Explanation (Optional):
A. Incorrect: At 100% power Nitrogen is aligned to drywell pneumatics. The Nitrogen Storage tank is the normal source of nitrogen and maintains the supply header at - 110 psig. This alarm means that pressure from the Nitrogen Storage tank has lowered and the Backup nitrogen bottles are now in service.
B. Correct: At 100% power Nitrogen is aligned to drywell pneumatics. The Nitrogen Storage tank is the normal source of nitrogen and maintains the supply header at - 110 psig. The backup nitrogen system is comprised of two sets of ten nitrogen cylinders.
One bank is set to maintain the backup supply header at approximately 100 psig. The other bank will supply the header when the pressure drops to approximately 94 psig. If the instrument supply header pressure drops to 102 psig, annunciator "AIR/N2 TO DRYWELL TROUBLE" (C904LC-F3) will alarm and the instrument supply header will be PNPS 2014 NRC Written 022014.doc
supplied by the backup nitrogen system. Instrument supply header pressure is indicated by Pl-4348, on Panel C-7 C. Incorrect: At 100% power, the drywell is inerted. When inerted, instrument air is valved out of service and manual action would be required to place it back in service.
D. Incorrect: These indications mean that pneumatic supply has swapped to the backup N2 bottles. Plausible in that there is a backup supply for extended SRV control that must be valves in.
Technical Reference(s): PNPS 2.2.1 05 page 6 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-08-02, E0-3.0 (As available)
Question Source: Bank # 6049 Minor editorial changes Modified Bank#
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295025 EK2.10 Importance Rating 2.9 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:
SPDS/ERIS/CRIDS/GDS: Plant-Specific Proposed Question: RO Question # 44 Regarding the SPDS system status windows shown below:
This indication means that ...
A. Torus water level is above the Tech Spec minimum.
B. RPV pressure is above the shutoff head of the LPCI pumps.
C. Cooling water is not available to the RHR pumps seal coolers.
D. RPV pressure is high outside the allowable validation range and is suspect.
Proposed Answer: B Explanation (Optional):
A. Incorrect: "WTR AVAILABLE" means that Torus water level is above the minimum required for pump operation(- 35 inches).
B. Correct: lAW PNPS 2.6.1, page 65, RPV PR HI means that the RPV pressure is higher than the shutoff head pressure of the system pump.
C. Incorrect: This means that pressure is above the shutoff head of the pump. Plausible in that the SOC Status is displaying "CLG NOT AVAIL". However this means that there is PNPS 2014 NRC Written 022014.doc
no Cooling Water flow to the RHR HX or the RPV temperature is not higher than the Cooling Water inlet temperature to the RHR HXs (see page 69 of PNPS 2.6.1)
D. Incorrect: This means that pressure is above the shutoff head of the pump. Plausible in that a validation check is performed on status windows based on the status of sensors.
Technical Reference(s): PNPS 2.6.1, page 65, (Attach if not previously provided)
DETERMINE PLANT SYSTEM STATUS USING EPIC-SPDS.
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-11-01, Terminal Objective, (As available)
DETERMINE PLANT SYSTEM STATUS USING EPIC-SPDS.
Question Source: Bank#
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295026 EK3.04 Importance Rating 3. 7 EK3.04 - Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: SBLC injection Proposed Question: RO Question # 45 During ATWS conditions, reactor power is 10% following the tripping of the Recirc Pumps.
Given the above, what are the bases for injecting Boron before torus water temperature exceeds the Boron Injection Initiation Temperature?
A. It ensures that the Cold Shutdown Boron Weight will be injected before the Torus temperature reaches the Heat Capacity Temperature Limit.
B. It ensures that the Hot Shutdown Boron Weight will be injected before the Torus temperature reaches the Heat Capacity Temperature Limit.
C. It ensures that the Cold Shutdown Boron Weight will be injected before the Torus bottom pressure reaches the Primary Containment Pressure Limit.
D. It ensures that the Hot Shutdown Boron Weight will be injected before the Torus bottom pressure reaches the Primary Containment Pressure Limit.
Proposed Answer: B Explanation (Optional):
A. Incorrect: The curve is based on injecting the Hot Shutdown Boron weight.
B. Correct: THE BIIT curve is defined as the greater of either:
- 1) The torus water temperature at which a reactor scram is required by plant Technical Specifications ( 11 0°F) or
- 2) The highest torus water temperature at which initiation of boron injection using SBLC will result in injection of the Hot Shutdown Boron Weight (HSBW) before torus water temperature exceeds the Heat Capacity Temperature Limit (HCTL).
For reactor power levels below 12.5%, if boron is injected before torus water PNPS 2014 NRC Written 022014.doc
temperature reaches BIIT, the torus temperature will remain below the HCTL during RPV blowdown. This ensures that all of the steam will be suppressed in the torus, and will minimize the pressure increase of the containment during a blowdown.
C. Incorrect: The limit of concern is the HCTL, not the PCPL. Plausible in that one of the objectives of EOP-02 is to protect the primary containment by lowering water level when required. Additionally the curve is based on injecting the Hot Shutdown Boron weight.
D. Incorrect: The limit of concern is the HCTL, not the PCPL. Plausible in that one of the objectives of EOP-02 is to protect the primary containment by lowering water level when required.
Technical Reference(s): 0-R0-03-04-02, Page 77 and 78 (Attach if not previously provided) 0-R0-03-04-04, Page 61 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-04, EO 20 (As available)
Question Source: Bank # WTSI295026 Modified Bank #
New Question History: Last NRC Exam: 2009 Hope Creek Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295021 AK3.01 Importance Rating 3.3 Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING : Raising reactor water level Proposed Question: RO Question # 46 Given the following conditions:
- The plant is depressurized with coolant temperature at 150 degrees F.
- M0-1001-50 Valve (Shutdown Cooling Suction) fails closed and cannot be opened by any means.
- Both Reactor Recirc Pumps are tagged out.
Under these conditions, PNPS 2.4.25, "Loss of Shutdown Cooling" requires reactor water level be _ _
A. maintained below the Group I isolation setpoint to ensure that Bypass Valves are available for use in the event that the plant becomes pressurized.
B. maintained below the HPCI Hi Level trip point to ensure that HPCI is available for use in the event that the plant becomes pressurized.
C. raised above +60 inches to promote natural circulation.
D. raised above +60 inches in preparation for initiating cooling by feed and bleed.
Proposed Answer: C Explanation (Optional):
A. Incorrect- must be maintained> +60 inches B. Incorrect- must be maintained > +60 inches C. Correct - Per PNPS 2.4.25, Loss of Shutdown Cooling if forced the reactor is NOT pressurized or if no heat sink is available, you must raise level above +60 inches in order to promote natural circulation or start a recirc pump to promote natural circulation.
PNPS 2014 NRC Written 022014.doc
D. Incorrect- feed and bleed is performed with the reactor pressurized Technical Reference(s): PNPS 2.4.25 Step 4.0[6]b. (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-01-18, TO DIRECT (As available)
CORRECTIVE ACTIONS TO MITIGATE CONSEQUENCES OF ABNORMAL EVENT.
Question Source: Bank# X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: 20011 Pilgrim, #55 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2, 10 55.43 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295024 EK3.05 Importance Rating 3.5 EK3.05- Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE : RPV flooding Proposed Question: RO Question# 47 A leak in the drywell has resulted in high drywell pressure and temperature. Emergency RPV Depressurization has been conducted in response to the containment parameters. Current plant conditions are as follows:
- Drywell pressure: 40 psig rising
- Torus Bottom pressure: 38 psig rising
- Drywell temperature: 340 °F, rising
- Reactor Pressure: 60 psig, lowering slowly
- All control rods are inserted
- Reactor level:
o oscillating full scale on all Narrow Range instruments o oscillating between the bottom of scale and -80 inches on the Fuel Zone instruments o Shutdown and FWLC instruments are pegged high.
Which one of the following is required and why?
A. Primary Containment Flooding in order to prevent containment failure.
B. RPV Flooding in order to establish and maintain adequate core cooling via steam cooling.
C. RPV Flooding in order to establish and maintain adequate core cooling via core submergence.
D. Primary Containment Flooding in order to establish and maintain adequate core cooling via core submergence.
Proposed Answer: C PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: RPV Flooding is required because RPV level cannot be determined.
Plausible due to the very high containment parameters and that the RPV has already been depressurized.
B. Incorrect: RPV Flooding floods the RPV to the main steam lines and establishes core cooling via core submergence.
C. Correct: The high drywell temperature and low RPV pressure have resulted in plant conditions exceeding the RPV Saturation Temperature limit which may result in instrument run flashing. The indications provided are indicative that flashing is occurring and RPV level cannot be determined. RPV flooding is then required. RPV Flooding floods the RPV to the main steam lines and establishes core cooling via core submergence.
D. Incorrect: RPV Flooding is required. Plausible in that Primary Containment Flooding is required if RPV water level cannot be restored and maintained above -150 inches and the Fuel Zones are oscillating below -150 inches.
Technical Reference(s): 0 -R0-03-04-08, page 7 (Attach if not previously provided)
EOP Caution 1 EOP-01, RPV Control, Overrides L-2 and P-1 Proposed References to be provided to applicants during examination: None Learning Objective: 0 -R0-03-04-08, EO 1 and 3 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load PNPS 2014 NRC Written 022014.doc
changes, and operating limitations and reasons for these operating characteristics.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295037 EA 1.07 Importance Rating 3.9 Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : RMCS: Plant-Specific Proposed Question: RO Question # 48 PNPS 5.3.23, Alternate Rod Insertion, is being used to manually insert control rods during ATWS conditions.
The Rod Worth Minimizer has just been bypassed and drive water pressure has been increased to 400 psid.
lAW 5.3.23, which one of the following correctly describes how control rods are now inserted?
First insert ...
A. the RPR array using the Emergency In I Notch Override Switch.
B. the RPR array using the continuous insert feature of the Rod Movement Control Switch.
C. control rods to form a checkerboard in a spiral pattern from the outside to the center using the Emergency In I Notch override Switch in the EMER IN position.
D. control rods to form a checkerboard in a spiral pattern from the outside to the center using the continuous insert feature of the Rod Movement Control Switch.
Proposed Answer: A Explanation (Optional):
A. Correct: The RPR array is inserted first using the Emergency In Switch.
B. Incorrect: 5.3.23 specifies that the Emergency In switch be used. Plausible in that placing and holding the Rod Movement Control Switch will cause the rod to continually insert (vice inserting just one notch).
C. Incorrect: The RPR array is inserted first. Plausible in that after the RPR array is inserted then the rods are inserted to form a checkerboard pattern.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: The RPR array is inserted first. Plausible in that after the RPR array is inserted then the rods are inserted to form a checkerboard pattern. Additionally, use of the Emergency In switch is specified.
Technical Reference(s): PNPS 5.3.23, Section 3.1, step (Attach if not previously provided)
[3](e)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-04, TO IMPLEMENT (As available)
REQUIRED CORRECTIVE ACTIONS FORANATWS Question Source: Bank#
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295016 AA1.01 Importance Rating 3.8 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT : RPS Proposed Question: RO Question # 49 The Main Control Room has been evacuated due to a fire. No operator actions were taken before the control room was evacuated.
lAW PNPS 2.4.143, Shutdown From Outside the Control Room
( 1) What method is used to insert a reactor scram AND (2) Assuming that EPIC is not available what method is used to determine reactor shutdown status?
A. ( 1) Manually venting the scram air header locally (2) Verifying the scram air header pressure has decayed to zero B. (1) Manually venting the scram air header locally (2) Verifying that the scram inlet and outlet valves have opened on all HCUs C. (1) Opening the Reactor Protection System Logic Channel "A" and "B" breakers on Panel C511 (2) Verifying the scram air header pressure has decayed to zero D. (1) Opening the Reactor Protection System Logic Channel "A" and "B" breakers on Panel C511 (2) Verifying that the scram inlet and outlet valves have opened on all HCUs Proposed Answer: D Explanation (Optional):
A. Incorrect: The method prescribed in PNPS 2.4.143 is to open the breakers on C511.
Additionally since EPIC is not available the only method remaining to determine PNPS 2014 NRC Written 022014.doc
shutdown status is by verifying the status of the scram inlet and out let valves on the HCUs.
B. Incorrect: The method prescribed in PNPS 2.4.143 is to open the breakers on C511.
C. Incorrect: Since EPIC is not available the only method remaining to determine shutdown status is by verifying the status of the scram inlet and out let valves on the HCUs.
Plausible in that venting the air header is a method prescribed in PNPS 5.3.23 dueing ATWS conditions.
D. Correct: PNPS 2.4.143 directs that if possible, the reactor be scrammed before the control room is evacuated. If not possible then the procedure via step [16](a) directs that Reactor Protection System Logic Channel "A" and "B" breakers on Panel C511 be opened which will de-energize RPS and should result in control rod insertion.
Step [16](g) then directs that the shutdown status be determined. With EPIC not available the only method available listed in Appendix M and as discussed in Discussion Step 5.0[9] is verifying that scram inlet and outlet valves have opened.
Technical Reference(s): PNPS 2.4.143, pages 10 and 12 (Attach if not previously provided)
PNPNS 2.4.143 Discussion Step 5.0[9] and Appendix M Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank #
Modified Bank# PNPS 12691 Modified to include method for determining shutdown status.
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295003 AA 1. 04 Importance Rating 3.6 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER: D.C. electrical distribution system Proposed Question: RO Question # 50 The plant is at rated conditions with all equipment operable when the following alarms are received on panel C3:
- AUX/SU XFMR A-6 SUPPLY TRIP (C3LC-E2)
The reactor is about to manually scrammed.
Assuming no other operator action is taken, the status of the 4KV Emergency Buses one minute after the scram will be:
A A-5 energized by the Startup Transformer A-6 energized by the Startup Transformer B. A-5 energized by the Startup Transformer A-6 energized by the Shutdown Transformer C. A-5 energized by the Startup Transformer A-6 de-energized D. A-5 energized by the Shutdown Transformer A-6 de-energized Proposed Answer: C Explanation (Optional):
A Incorrect: A-6 is de-energized. Plausible if the candidate does not recognize the symptoms of a D-17 loss. Alternatively the candidate may not realize the impact on the Startup Transformer supply breaker following the loss of D-17 (loss of control power).
PNPS 2014 NRC Written 022014.doc
B. Incorrect: A-6 is de-energized. Plausible if the candidate does not recognize that the Shutdown Transformer supply is also impacted by the DC loss.
C. Correct: The alarms are symptomatic of a loss of 125 VDC bus D-17. This will result in a complete loss of control power to bus A6. When the reactor is scrammed and the turbine trips bus A6 will de-energize. Bus A5 which still has control power, will fast transfer to the Startup Transformer.
D. Incorrect: Bus A5 will fast transfer to the Startup Transformer. Plausible if the candidate believes that the Startup Transformer transfer logic is disabled by the loss of D-17.
Technical Reference(s): PNPS 5.3.12, Attachment 1, (Attach if not previously provided)
Caution 3.
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-01-02 EO 8 (As available)
Question Source: Bank# X Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: Not Used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295038 EA2.04 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : Source of off-site release Proposed Question: RO Question# 51 The plant is operating at full power with the AOG system aligned for normal full power operation. The following indications are observed:
- Pre Treat Rad Monitors are steady
- Post Treat Rad Monitors are increasing
- Main Stack Rad Monitors are increasing
- Normal Recombiner temperatures for full power operation
- Alarm ADSORBR VESSEL TEMP HI, CP600R-B2, alarms Which one of the following could result in these indications?
A. Low steam dilution flow B. High steam dilution flow C. Wetting of the Recombiner D. Wetting of the Charcoal Adsorbers Proposed Answer: D Explanation (Optional):
A. Incorrect: Low steam dilution flow would result in higher than normal recombiner temperatures.
B. Incorrect: High dilution flow would result in abnormal recombiner temperature.
C. Incorrect: This would result in lower recombiner temperatures.
D. Correct: lAW PNPS 2.2.1 06, Attachment 9, item H, charcoal performance will deteriorate when the charcoal gets wet. Holdup times for krypton and xenon will PNPS 2014 NRC Written 022014.doc
decrease and plant emissions will increase. The Post Treat Monitors and the Mani Stack Monitors are both downstream of the Adsorbers.
Technical Reference(s): PNPS 2.2.1 06, Attachment 9, item (Attach if not previously provided)
H PNPS 2.4.136, Section 5.0, discussion items [1] and [2]
PNPS 2.4.141, Section 5.0, discussion item [1]
Proposed References to be provided to applicants during examination: None Learning Objective: O-R0-02-04-11, EO 3k. 9d (As available)
Question Source: Bank # Made editorial and format changes.
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 700000 AA2.04 Importance Rating 3.6 Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: VARs outside capability curve.
Proposed Question: RO Question # 52 The plant is at power with the following initial Main Generator Parameters:
- MWe loading: 680
- MVAR loading: 100 lagging
- Hydrogen Gas pressure: 45 psig
- Main Generator Voltage Regulation: Auto Then, a disturbance on the grid causes grid voltage to slowly lower over time.
Given the expected response of the Main generator to the above, when will the FIRST Main Generator Limit be exceeded? Note: The Generator Capability Curve is provided for your use.
When MVARs A. rise to 170 MVARs lagging B. rise to 360 MVARs lagging C. lower to 100 MVARs leading D. lower to 180 MVARs leading Proposed Answer: A Explanation (Optional):
A. Correct: As grid voltage lowers the voltage regulator will try to maintain its generator output voltage setpoint by raising field current. This will result in the generator picking up reactive system loads. This will cause the MVAR loading to increase in the lagging direction. The 45 psig hydrogen pressure curve will be exceeded when MVAR loading exceeds 170 MVARs.
PNPS 2014 NRC Written 022014.doc
B. Incorrect: A limit was first exceeded when MVAR loading exceeds 160 MVARs.
C. Incorrect: MVARs will increase in the lagging direction as the generator picks up reactive loading as grid voltage lowers. This is plausible if the candidate believes the opposite and applies the 100 MVAR leading limit of PNPS 2.2.2 (100 MVAR leading).
D. Incorrect: MVARs will increase in the lagging direction This is plausible if the candidate believes the opposite and using the 45 psig hydrogen pressure limit. If so the limit will be exceeded when MVARs lower to 160, leading.
Technical Reference(s): 2.4.144, Attachment 4 (Attach if not previously provided) 2.2.2 Precautions 4 and 5 MOTORS AND GENERA TORS LP Proposed References to be provided to applicants during examination: Attachment 4 of PNPS 2.4.144.
Learning Objective: 0-R0-02-01-08, E017f (As available)
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295030 EA2.01 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool level Proposed Question: RO Question# 53 Which ONE of the following corresponds to the maximum torus water level at which the HPCI System must be secured and injection prevented irrespective of adequate core cooling?
A. 95 inches B. 90 inches C. 85 inches D. 50 inches Proposed Answer: A Explanation (Optional}:
A. Correct: lAW EOP-3, HPCI must be secured at 95" because the turbine exhaust will be above the water level in the torus and this can lead to over pressurizing the primary containment.
B. Incorrect: lAW- EOP-3, this torus level uncovers the Downcomers and require E-Depress.
C. Incorrect: This value is too low and will result in HPCI pressurizing the primary containment.
D. Incorrect: This is the value associated with uncovering the SRV tail pipes.
Technical Reference(s): EOP-3, Steps TL-12 & 13 (Attach if not previously provided)
PNPS 2014 NRC Written 022014.doc
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # WTSI # 1459 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Shielding, isolation, and containment design features, including access limitations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295018 2.4.47 Importance Rating 4.2 Emergency Procedures I Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (Partial or Total Loss of CCW)
Proposed Question: RO Question # 54 The plant is at full power with a normal configuration on the Salt Service Water System.
The control room crew notes the following changes in SSW and RBCCW indications:
Using the partial SSW P&ID provided, which one of the following is consistent with the above changes in SSW parameters?
A. "B" Loop SSW pump impeller failure B. Fouling of the "B" RBCCW heat exchanger C. Fouling of the "B" TBCCW heat exchanger D. Leak on the outlet of the "A" RBCCW heat exchanger Proposed Answer: B Explanation (Optional):
A. Incorrect: The discharge of all SSW pumps is tied together. An impeller failure would reduce the pressure and flow in both loops.
B. Correct: Fouling of the "B" RBCCW HX would reduce the flow in the "B" loop. This would cause the flow in the "A" loop to go up. Since the loops are tied together the pressure would be the same in both loops.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: Fouling of the "B" TBCCW HX would cause the flow to go up in both loops because the flow indicators are on the inlets of the two RBCCW HXs (see drawing).
D. Incorrect: A leak on the outlet of the "A" RBCCW HX would cause flow in the "A" loop to go up as shown because the flow indicator is on the inlet to the HX. This would also cause "B" loop flow to go down. However pressure in the system would lower not rise.
Technical Reference(s): P&ID M212 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: Partial SSW P&ID, showing just HX's and CR indications.
White-out anything doing with automatic actions such as Load shed on SSW valves.
Learning Objective: 0-R0-02-02-02, Terminal Objective, (As available)
Respond to a loss of one SSW loop.
Question Source: Bank # LOR Bank Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295005 2.4.21 Importance Rating 4.0 Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (Main Turbine Generator Trip)
Proposed Question: RO Question # 55 Following a Main Turbine Trip from rated conditions EPIC displays the following indication:
- CALLRODS indicates "YES" The CALLRODS indication means that:
A. All control rods are at least at position 02.
B. The program has activated but has not yet determined rod status.
C. No more than one rod is fully withdrawn and all other control rods are at least at position 04.
D. No more than one control rod is fully withdrawn and all other control rods are fully inserted.
Proposed Answer: A Explanation (Optional):
A. Correct: Upon receipt of a Scram signal, the CALLRODS program is triggered by backup Scram valve relay 5A-K21 B. After a 25-second time delay to allow the control rods to settle, the program evaluates control rod position by evaluating whether the 00 or 02 position switch is currently made up. Any control rods which do not satisfy this criteria are stored for further evaluation in remaining phases. If all 145 control rods satisfy the criteria, the CALLRODS point changes status to NML YES. Although there are other phases to the program all that can be said following a "YES" indication is that all rods are at least at position 02.
B. Incorrect: The YES indication means that the program has completed its run and determines that all rods are to or beyond position 02. Plausible if the candidate is not PNPS 2014 NRC Written 022014.doc
familiar with the program nomenclature but understands that the program may take up to 3 minutes to determine rod status.
C. Incorrect: This condition would result in a "NO" indication. Plausible in that one rod being fully withdrawn would satisfy TS shutdown margin and position 04 is the Maximum Subcritical Banked Withdrawal Position.
D. Incorrect: This condition would satisfy shutdown margin.
Technical Reference(s): PNPS 2.2.89, Discussion item [6] (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-04, E0-4.a, 4.b (As available)
Question Source: Bank # PNPS 3133 Modified Bank #
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295006 2.2.42 Importance Rating 3.9 Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical Specifications (SCRAM).
Proposed Question: RO Question # 56 Which one of the following instrument failures would require an Active LCO as specified in Tech Specs and PNPS 1.3.34.2, Limiting Conditions for Operation Log?
Assume all other instruments are operable.
A. IRMs "A" and "C" are inoperable with reactor power at 100%.
B. APRM "A" is inoperable with the reactor critical at 200 cps in the source range.
C. One reactor high pressure scram switch is inoperable with the Reactor Mode Switch in Startup.
D. One SDIV high level scram switch is inoperable with the Reactor Mode Switch in Shutdown.
Proposed Answer: C Explanation (Optional):
A. Incorrect: IRM's are bypassed and not required when APRM's are on scale and the reactor mode switch is in the Run position. Plausible in that 3 of the 4 IRMs per trip system are required when the IRMs are required to be operable. This would warrant a Tracking LCO per PNPS 1.3.34.2.
B. Incorrect: APRMs are required to be operable with the Mode Switch in Startup. With the reactor critical the Mode Switch is in Startup. However only 2 of the 3 available APRMs per trip system are required. This would warrant a Tracking LCO per PNPS 1.3.34.2.
C. Correct: An active LCO is required when the inoperable component or system is required by the Technical Specifications for the current plant condition. Per TS Table 3.1.1, all reactor high pressure scram switches are required to be operable with the Mode Switch in Startup.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: No scram functions are required when the Mode Switch is in Shutdown.
Plausible in that this function is required in all mode switch positions other than Shutdown with some caveats while in Refuel. This would warrant a Tracking LCO per PNPS 1.3.34.2.
Technical Reference(s): PNPS 1.3.34.2 section 3.0, items (Attach if not previously provided)
[1] and [12]
TS Table 3.1.1 and associated notes Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-07 -07, EO 23 (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295031 EA2.03 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL : Reactor pressure Proposed Question: RO Question# 57 Following a trip of "A" Recirc Pump, the plant is operating with the "B" Recirc Pump in service.
Then a complete loss of feed occurs. Plant conditions are as follows:
- Current Drywell Pressure: 1.4 psig, slowly lowering
- Current RPV level: -30 inches, slowly rising
- HPCI has just been placed in pressure control
- Current RPV Pressure: 945 psig and lowering slowly
- Lowest RPV level during transient: - 80 inches
- M0-1001-28A, LPCIINJ THROTTLE VLV#1: Open
- M0-1001-29A, LPCIINJ VLV #2: Closed
- M0-1001-288, LPCIINJ THROTTLE VLV#1: Open
- M0-1001-298, LPCIINJ VLV #2: Closed Which one of the following statements is correct regarding LPCI Loop Select? Assume no operator actions have been taken relative to RHRILPCI.
LPCI Loop select ...
A. should have closed M0-1001-28A B. should have closed M0-1001-288 C. will close M0-1001-28A during the subsequent plant cooldown D. will close M0-1001-288 during the subsequent plant cooldown Proposed Answer: C PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect- LPCI Loop select activated when water level lowered below -46 inches.
Normally it will select a loop shortly after activating by examining the Dp across the jet pump risers. If it cannot discern a sufficient difference it will default to loop "B". In single loop however (or no pumps running), it will wait till RPV pressure lowers below 900 psig to select a loop. Since pressure is 945 psig this has not yet occurred. The closure signal will be sent to 28A.
B. Incorrect- LPCI Loop select activated when water level lowered below -46 inches.
Normally it will select a loop shortly after activating by examining the Dp across the jet pump risers. If it cannot discern a sufficient difference it will default to loop "B". In single loop however (or no pumps running), it will wait till RPV pressure lowers below 900 psig to select a loop. Since pressure is 945 psig this has not yet occurred.
C. Correct- LPCI Loop select activated when water level lowered below -46 inches.
Normally it will select a loop shortly after activating by examining the Dp across the jet pump risers. If it cannot discern a sufficient difference it will default to loop "B". In single loop however (or no pumps running), it will wait till RPV pressure lowers below 900 psig to select a loop. Since pressure is 945 psig this has not yet occurred. When pressure lowers to 900 psig it will select a loop. Because the transient was initiated by a loss of feed the riser dps will be the same and the "B' loop will be selected. This will send a closure signal to M0-1 001-28A.
D. Incorrect- Because the transient was initiated by a loss of feed the riser dp's will be the same and the "B' loop will be selected. This will send a closure signal to M0-1001-28A.
Technical Reference(s): 2.2.19, Attachment 9 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # LOR Bank #17 Modified Bank#
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge PNPS 2014 NRC Written 022014.doc
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295001 2.2.25 Importance Rating 3.2 Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (Partial or Complete Loss of Forced Core Flow Circulation)
Proposed Question: RO Question # 58 The plant is at rated conditions.
Then a malfunctioning "B" Recirc Flow controller results in a lowering "B" Recirc pump speed.
Additional information is as follows:
- The reactor operator inserts a scoop tube lock on the "B" Recirc MG set and the speed reduction is terminated.
- Reactor power stabilizes at 85%.
- "B" Recirc pump speed stabilizes at 68%
- "A" Recirc pump speed is 80%
(1) Has the Tech Spec Recirc Speed Mismatch LCO been exceeded and (2) What are the bases for this TS limitation?
A. (1) Yes (2) Minimize Jet Pump Vibration B. ( 1) No (2) Minimize Jet Pump Vibration C. (1) Yes (2) Ensure proper functioning of the LPCI Loop Selection logic D. ( 1) No (2) Ensure proper functioning of the LPCI Loop Selection logic Proposed Answer: C Explanation (Optional):
A. Incorrect: The limit is based on ensuring the proper functioning of the LPCI Loop PNPS 2014 NRC Written 022014.doc
Selection logic.
B. Incorrect: The limit is a 10% speed mismatch when power is above 80%. Therefore the limit was exceeded when pump speed lowered below 70%. Plausible if the candidate applies the limit of 15% which is only applicable below 80% power. Additionally, the limit is based on ensuring the proper functioning of the LPCI Loop Selection logic.
There is a concern with excessive vibration but this is addressed by the minimum pump speed limits which protect against cavitation (see PNPS 2.2.84, Precaution 4).
C. Correct: The limit is a 10% speed mismatch when power is above 80%. Therefore the limit was exceeded when pump speed lowered below 70%. As discussed in the TS bases forTS 3.6.F.1 the licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%. At or below 80% power the loop select logic would not be expected to function at a speed differential of 20%. This specification provides a margin of 5% in pump speed differential before a problem could arise.
D. Incorrect: The limit is a 10% speed mismatch when power is above 80%. Therefore the limit was exceeded when pump speed lowered below 70%.
Technical Reference(s): TS 3.6.F.1 and associated bases. (Attach if not previously provided)
PNPS 2.2.84, Precaution 4 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-06-02 EO 35f (As available)
Question Source: Bank # X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, PNPS 2014 NRC Written 022014.doc
effects of load changes, and operating limitations and reasons for these operating characteristics.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295014 AK1.01 Importance Rating 3. 7 Knowledge of the operational implications of the following concepts as they apply to INADVERTENT REACTIVITY ADDITION : Prompt critical Proposed Question: RO Question # 59 Control rods are being withdrawn for a reactor plant startup.
- The Rod Worth Minimizer is bypassed
- A second operator is verifying rod movement but several control rods are withdrawn out of sequence
- The reactor has just been taken critical
- Reactor power is 1000 CPS in the source range Then, an uncoupled, fully inserted, center control rod blade falls to the full out position.
All other systems respond as designed.
Regarding the fuel bundles around the dropped rod, the transient will result in .....
A. no cladding or fuel damage B. only minor cladding perforations C. instantaneous fragmentation and dispersal of the fuel rods D. complete melting of the fuel pellets but no significant cladding failure Proposed Answer: C Explanation (Optional):
A. Incorrect: Plausible if the operator believes that the combination of the control rod velocity limiter, fuel loading and the Reactor Protection System will prevent damage.
B. Incorrect: Plausible in that per the Safety Analysis, if control rods are withdrawn in-sequence and a control rod drop accident should occur, the highest attainable heat generation is <220 cal./gm. Conservatively, this is estimated to result in perforating 330 PNPS 2014 NRC Written 022014.doc
fuel rods. However in this case multiple control rod errors have occurred.
C. Correct: The worst case rod drop accident would be one which occurred in the startup region. Multiple operator withdraw and insert errors could establish a critical central core region with the center rod fully inserted If this rod were then to suffer a rod drop accident, considerable fuel damage could result when the reactor goes prompt critical.
Enthalpies greater than 425 cal/gram may result. This would produce instantaneous fragmentation and dispersal of the fuel rods. Pressure would increase explosively, and fission fragment release from the clad would be gross.
D. Incorrect: Plausible in that 280 cal/gram is the design limit for the Rod drop accident.
Although complete U02 melting does occur if the fuel enthalpy reaches 280 cal/gram, relatively few fission products are released from the clad and the resulting pressure transient is minimal. This is the bases for the Tech Spec limit.
Technical Reference(s): Rod Worth Minimizer reference (Attach if not previously provided) text, Attachment 1 (pages 29 & 30)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: NA Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295033 EK2.02 Importance Rating 3.8 Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Process radiation monitoring system Proposed Question: RO Question # 60 Which one of the following alarms requires immediate entry into EOP-04?
A. SBGT DISCH RAD HI, C904LC-F4 B. REACTOR BLDG RAD HI, C904LC-A1 C. REFUEL FLOOR RAD HI, C904LC-C7 D. REACTOR BLDG VENT RAD HI C904LC-B5 Proposed Answer: D Explanation (Optional):
A. Incorrect: This alarm in and of itself, does not require entry or re-entry into EOP-04.
Plausible in that this alarm may occur if SBGT was in service while a primary system was discharging into the secondary containment.
B. Incorrect: EOP-04 is entered on high area radiation levels as determined by local survey lAW PNPS 5.3.33, SECONDARY CONTAINMENT RADIATION SURVEYS FOR EOP-04. The entry condition does not utilize Hi Rad alarms associated with the Rx.
Building.
C. Incorrect: A Refuel Floor High radiation condition is not addressed by EOP-04.
D. Correct: The EOP-04 entry is required if annunciator REACTOR BLDG VENT RAD HI C904LC-B5 alarms. This is because the alarm Indicates a release of radioactivity to the Reactor Building environment.
Technical Reference(s): EOP-04, Entry Conditions (Attach if not previously provided)
PNPS 2014 NRC Written 022014.doc
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-06, EO 1 (As available)
Question Source: Bank # LOR Bank Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295029 EK3.01 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL : Emergency depressurization Proposed Question: RO Question# 61 EOP-03, Primary Containment Control requires that an Emergency Depressurization be performed when torus level cannot be restored and maintained below _ ( 1) _ _.
This is because_ (2) _ .
A. (1) 175inches (2) the torus to drywell vacuum breakers begin to cover at this value, increasing the likelihood of exceeding the negative design pressure of the drywell.
B. (1) 175inches (2) torus level has reached the SRV Tail Pipe Level Limit (STPLL)
C. (1) 300 inches (2) the torus to drywell vacuum breakers begin to cover at this value, increasing the likelihood of exceeding the negative design pressure of the drywell.
D. (1) 300 inches (2) the torus level has reached the SRV Tail Pipe Level Limit (STPLL)
Proposed Answer: B Explanation (Optional):
A. Incorrect: The reason is because the STPLL is being exceeded. See below. Plausible in that there is an action in EOP-03 that addresses covering up the torus to drywell vacuum breakers at 180 inches.
B. Correct: Step TL-7, directs that when torus level cannot be restored and maintained below 175 inches then an Emergency Depress is to be performed. This value corresponds to the SRV Tail Pipe Level Limit (STPLL). The STPLL is the lower of either:
(1) The Maximum Pressure Suppression Primary Containment Water Level PNPS 2014 NRC Written 022014.doc
(MPSPCWL, 175 in.) OR (2) The highest torus water level at which opening an SRV will not result in exceeding the allowable stresses in the SRV tailpipes, tailpipe supports, quenchers, or quencher supports.
The MPSPCWL (175 in.) is more limiting and defines the PNPS STPLL.
The STPLL (MPSPCWL) is therefore the water level corresponding to the bottom of the ring header in the torus. It is the upper torus water level bound that ensures pressure suppression capability of the primary containment during emergency depressurization through SRVs.
C. Incorrect: The limit for Emergency Depress is 175 inches. Plausible in that there is an action in EOP-03 that addresses covering up the torus to drywell vacuum breakers at 180 inches. If a candidate thinks that the vacuum breakers tap off the top of the torus, this would lead him/her to select this answer.
D. Incorrect: The limit for Emergency Depress is 175 inches.
Technical Reference(s): EOP-03 lesson plan, page 33 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-02, EO 27 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Shielding, isolation, and containment design features, including access limitations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295012 AA1.01 Importance Rating 3.5 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell ventilation system Proposed Question: RO Question # 62 The plant is at rated conditions with the drywell cooling fans aligned as follows:
- Four fans are in Standby
- All other fans are in Run Then a small steam leak results in rising drywell temperature and pressure and the reactor scrams. Five minutes later, plant conditions are as follows:
- Drywell Temperature: 220 degrees and rising
- Drywell Pressure: 4.5 psig and rising
- Startup Transformer: Locked Out
- Bus A5: Powered by "A" EDG
- Bus A6: Powered by "B" EDG
- No operator action has been taken regarding the drywell cooling system The drywell cooling fans that were originally in Run will be_ (1) __.
The drywell cooling fans that were originally in Standby will be _ (2) __.
A. ( 1) Running (2) Off B. ( 1) Off (2) Running C. (1) Running (2) Running D. ( 1) Off (2) Off Proposed Answer: D PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: None of the fans will be running. Plausible if the candidate believes that the fans originally in Run will restart when the EDGs re-energize the bus and since they restarted there is no need for the standby fans to start on low flow.
B. Incorrect: None of the fans will be running. Plausible if the candidate believes that the standby fans will auto start on low flow after the EDGs re-energize their busses after the running fans tripped on loss of power.
C. Incorrect: None of the fans will be running. Plausible if the candidate recalls that the standby fans auto start when the Unit Aux Feeder breakers trip open following a scram after a 45 second time delay.
D. Correct: The EDGs supplying the bus during a high drywall pressure condition will generate load shed signals and trip all the drywall cooling fans supplied by the EDGs and prevent any auto starts.
Technical Reference(s): PNPS 2.2.49, 9 (Attach if not previously provided)
PNPS 2.4.16, Attachment 8, sheet 2
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-08-04, EO 8c (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295034 EA2.02 Importance Rating 3. 7 EA2.02 - Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Cause of high radiation levels Proposed Question: RO Question # 63 The plant is at 2% power and starting up with Primary Containment inerting in progress lAW PNPS 2.2. 70, Primary Containment Atmosphere Control System.
Annunciator C904LC, REACTOR BLDG VENTRAD HI now alarms.
Which one of the following potential sources could result in this alarm?
A. Recirc Pump seal failure B. Reactor Feed pump seal failure C. Steam leak in the condenser compartment D. High activity on the discharge of the Mechanical Vacuum Pump Proposed Answer: C Explanation (Optional):
A. Incorrect: Although the drywell is being vented this would not result in an increase in REACTOR BLDG VENT rad levels because SBGT is used while inerting the containment. SBGT discharges to the main stack. Plausible if the candidate believes that inerting is done using the reactor building ventilation exhaust fans which is physically possible.
B. Incorrect: The RFP ventilation would exhaust the contaminants out the roof exhausters.
C. Correct: The Turbine Building Ventilation system supplies the condenser compartment.
This system discharges to the Reactor Building Vent via the Turbine Building Exhaust Fans. This would then be seen on the Rx Building Vent PRMs.
D. Incorrect the Mechanical Vacuum pump discharges to the Main Stack PNPS 2014 NRC Written 022014.doc
Technical Reference(s): Figure 4, Turbine Building (Attach if not previously provided)
Ventilation Reference Text Figure 1, Reactor Building Ventilation Reference Text Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295036 2.2.14 Importance Rating 3.9 Equipment Control: Knowledge of the process for controlling equipment configuration or status.
Secondary Containment High Sump/Area Water Level Proposed Question: RO Question # 64 A fuel damaging event has resulted in EOP entry. Plant conditions are as follows:
- ATWS conditions exist
- SBLC is not available
- Many control rods have not inserted and control rods are being inserted using the RMCS
- A leak has been reported on the discharge of the only available CRD pump
- The CRD Quad is being pumped down and water level in the quad is currently stable at 4 inches
- The event has just been upgraded to a General Emergency based on the extent of fuel damage Which of the following is correct regarding EOP-04, Secondary Containment Control actions?
A. Leave the CRD pump in service.
No other action is required.
B. Secure and isolate the leaking CRD pump.
No other action is required.
C. Leave the CRD pump in service.
Open the Rx Building Floor and Equipment Sump pump breakers.
D. Secure and isolate the leaking CRD pump.
Open the Rx Building Floor and Equipment Sump pump breakers.
Proposed Answer: C Explanation (Optional):
A. Incorrect: The Rx Building Floor and Equipment Sump pump breakers should be opened.
PNPS 2014 NRC Written 022014.doc
B. Incorrect: Per step SC-12, systems are not isolated if they are required to be operated by EOPS. Without SBLC, CRD is required to achieve shutdown status. Plausible in that otherwise, it would be isolated since the level in the quad is above the Max Normal Operating Value.
C. Correct: lAW the 4th override in EOP-04, if plant conditions indicate that substantial core damage is occurring, the operator is directed to isolate the Reactor Building Sump pumps. If the Rx building sumps could be contaminated, opening the breakers for the RX building sump pumps will preclude pumping highly contaminated water outside the secondary containment. Per step SC-12, systems are not isolated if they are required to be operated by EOPS. Without SBLC, CRD is required to achieve shutdown status.
D. Incorrect: Per step SC-12, systems are not isolated if they are required to be operated by EOPS. Without SBLC, CRD is required to achieve shutdown status. Plausible in that otherwise, it would be isolated since the level in the quad is above the Max Normal Operating Value.
Technical Reference(s): EOP-04, 41h override and step SC- (Attach if not previously provided) 12 EOP -04 LP pages 13, 17, 18 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-06 EO 3 and TO #7 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295015 AK1.03 Importance Rating 3.8 Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM: Reactivity effects Proposed Question: RO Question # 65 Following a reactor scram, the following conditions are present:
- Four (4) rods are at position 04
- One ( 1) control rod is at position 20
- All other control rods are fully inserted
- Reactor Power is on IRM range 1 and lowering
- No Boron has been injected Which one of the following is correct regarding reactor shutdown status and whether a plant cooldown is permitted?
A. The reactor is not currently shutdown. A plant cooldown is not permitted.
B. Current reactor shutdown status cannot be determined without reactor engineering support. A plant cool down is not permitted.
C. The reactor is currently shutdown and will remain shutdown under all conditions if a reactor plant cool down is conducted. A plant cooldown is permitted.
D. The reactor is currently shutdown but may restart if a reactor plant cool down is commenced. A plant cooldown is permitted, while monitoring for a reactor restart.
Proposed Answer: D Explanation (Optional):
A. Incorrect: As defined in PNPS 5.3.35 definition [40] the reactor is shutdown when subcritical with Reactor power below the heating range (on scale on IRM range 7 or below). Plausible if the candidate confuses the definition of shutdown and shutdown under all conditions. The reactor is not shutdown under all conditions because not all rods are inserted to position 04 or beyond.
PNPS 2014 NRC Written 022014.doc
B. Incorrect: Definitions for shutdown and shutdown under all conditions are provided in PNPS 5.3.35 which allow an operator to make this determination without RE support.
Plausible in that 5.3.35 does recognize that an RE could also make these determinations for conditions not specified in 5.3.35.
C. Incorrect: The reactor is not shutdown under all conditions because not all rods are inserted to position 04 or beyond. Although a cooldown can be commenced the operator must monitor for a restart due to the reactivity effects of the cooldown.
D. Correct: As defined in PNPS 5.3.35 definition [40] the reactor is shutdown when subcritical with Reactor power below the heating range (on scale on IRM range 7 or below). EOP-02 directs that if the reactor is shutdown and no boron has been injected then a reactor cooldown is to be conducted. If the reactor should subsequently go critical due to the effects of the cooldown, EOP-02 Override P-7 will terminate further cooldown.
Technical Reference(s): PNPS 5.3.35 (Attach if not previously provided)
EOP-02 steps P-6 thru P-8 EOP Lesson Plan, page 50 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-04, EO 16, (As available)
Question Source: Bank #
Modified Bank# LOR 310 Changed shutdown status which changed the answer New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.1.20 Importance Rating 4.6 Ability to interpret and execute procedure steps.
Proposed Question: RO Question# 66 A system operating procedure is being performed. Under what conditions can procedure steps be performed out of sequence?
A. When the steps are "bulleted".
B. When the steps are "lettered".
C. Only when supervisor concurrence is obtained.
D. Only when the procedure authorizes specific steps to be performed out of order.
Proposed Answer: A Explanation (Optional):
A. Correct: lAW EN-HU-106, steps identified as bulleted steps can be performed in any sequence.
B. Incorrect: Lettered steps must still be performed in sequence. Plausible in that the steps are not numbered.
C. Incorrect: Supervisors cannot authorize procedure steps to be performed out of order.
Plausible in that supervisors can provide concurrence that a step is not applicable.
D. Incorrect: Bulleted steps can also be performed in any order regardless of whether there is explicit guidance in the procedure that certain steps can be performed out of order.
Technical Reference(s): EN-HU-1 06, section 5.2.2, item [3] (Attach if not previously provided)
PNPS 2014 NRC Written 022014.doc
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01, EO 129 (As available)
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.1 .31 Importance Rating 4.6 Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.
Proposed Question: RO Question # 67 During a loss of coolant accident in the drywell, HPCI automatically starts.
The operator notes that RPV level does not appear to be rising in response. The HPCI lineup on Panel C903 has been verified. Current HPCI indications are as shown below.
Based on the above indications which one of the fo llowing is correct A. The leak rate from the vessel must be approximately 4000 gpm.
B. There is a leak on the HPCI discharge line.
C. The HPCI pump impeller has failed.
D. The HPCI controller has failed.
Proposed Answer: B PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: Pump discharge pressure is < than turbine inlet pressure so HPCI injection flow is not making it to the reactor. Plausible in that the flow on the HPCI controller is 4000 GPM.
B. Correct: HPCI flow is 4000 GPM as called for by the controller. However Pump discharge pressure is < than turbine inlet pressure so HPCI injection flow is not making it to the reactor. Since the water is not going to the reactor it must be going somewhere
-i.e., there must be a leak on the discharge line.
C. Incorrect: If this were true the controller would not be showing 4000 GPM since the flow indicator is downstream of the pump.
D. Incorrect: The controller is in Auto and set to control HPCI flow at 4000 GPM. The system is delivering 4000 GPM as indicated on the controller. Plausible if the candidate does not understand the indications provided by the controller.
Technical Reference(s): HPCI Reference Text Figure 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01 (As available)
Question Source: Bank # LOR Bank Modified Bank#
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.2.40 Importance Rating 3.4 Ability to apply technical specifications for a system.
Proposed Question: RO Question # 68 The plant is at rated conditions. Which of the following situations would render the "A" Salt Service Water System inoperable?
Situation 1: Average seawater temperature is 72 oF Situation 2: 'A' RBCCW Heat Exchanger SSW Outlet Valve, M0-3800 jammed full open Situation 3: 'A' TBCCW Heat Exchanger SSW Outlet Valve, M0-3801 jammed full open A. Situation 1 only B. Situation 3 only C. Situations 2 and 3 only D. Situations 1 and 2 only Proposed Answer: B Explanation (Optional):
A. Incorrect: The Ultimate Heat sink (UHS) is operable up to a maximum seawater temperature of 75 oF B. Correct: Per Tech Spec bases, a SSW subsystem is considered OPERABLE when it has an OPERABLE UHS, two OPERABLE pumps with associated controls and instrumentation and the following valves on that subsystem operable:
- 1. One TBCCW heat exchanger outlet valve unless the valve is throttled.
- 2. One RBCCW heat exchanger outlet valve unless the valve is open.
With 3801 jammed full open it cannot reposition to its accident position (10%) open.
C. Incorrect: M0-3800 is full open and in its accident condition and therefore does not render the loop inoperable.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: Neither the seawater temperature or M0-3800 being full open renders the loop inoperable (see above)
Technical Reference(s): Tech Spec Bases page B3/4.5, (Attach if not previously provided) pages 15 and 15a Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # X Modified Bank #
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.2.6 Importance Rating 3.0 Knowledge of the process for making changes to procedures.
Proposed Question: RO Question# 69 While studying for an exam you discover a procedure error in the RHR system operating procedure.
In one section of the procedure, the RHR Loop "A" Torus Cooling/Spray Block Valve is identified as M0-1 001-34B.
The procedure owner in Merlin is listed as "ODM" and approval authority has not been officially assigned to designees within the organization.
The procedure change generated to fix the error would be classified as a/an_ (1) _ _
(intent /non-intent).
The lowest level person in the Operations Department that can approve the change is
_(2)_.
A. (1) intent (2) any SRO qualified shift manager B. ( 1) intent (2) the Operations Department Manager C. (1) non-intent (2) any SRO qualified shift manager D. ( 1) non-intent (2) the Operations Department Manager Proposed Answer: D Explanation (Optional):
PNPS 2014 NRC Written 022014.doc
A. Incorrect: lAW Attachment 6 of NOP 98A1, this change would be classified as a non-intent change. Plausible if the candidate believes non-intent changes are not allowed for safety related procedures. Plausible in that some types of changes for certain procedures are not allowed. For example EWN changes or Temporary Changes are not allowed for some 5.3 series procedures.
Additionally the change must be approved by the Operations Department Manager (ODM). Plausible in that a shift manager can approve an EWN change. However this change would not be associated with emergent work. If the candidate confuses the approval requirements, they would select this response.
B. Incorrect: lAW Attachment 6 of NOP 98A 1, this change would be classified as a non-intent change.
C. Incorrect: The change must be approved by the Operations Department Manager (ODM).
D. Correct: lAW Attachment 6 of NOP 98A 1, this change would be classified as a non-intent change which states that one criteria for a non-intent change is "Changing equipment identifiers (for example: changing "B" pump to "A" pump when the Procedure section obviously is written for the "A" pump). All other criteria are also satisfied.
lAW section 6.1.3.1 of NOP 98A 1, the procedure owner must approve all non-intent changes. In this case the ODM is the procedure owner.
Technical Reference(s): NOP98A1, Attachment 6 and (Attach if not previously provided) sections 5.3 and 6.1.3.1 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01, EO 2b (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X PNPS 2014 NRC Written 022014.doc
Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.3.4 Importance Rating 3.2 Knowledge of radiation exposure limits under normal or emergency conditions.
Proposed Question: RO Question # 70 Given the following:
- A Site Area Emergency has been declared due to a refueling accident.
- A re-entry team is being assembled to operate the bridge to move a bundle to its proper location.
- An individual assigned to the team has a lifetime accumulated dose to date of 4 Rem and an annual dose to date of 0.5 Rem.
- The Emergency Director has NOT authorized the use of the higher Emergency Exposure Limits of EP-IP-440 EMERGENCY EXPOSURE CONTROLS.
What is the MAXIMUM exposure this individual can receive without exceeding the requirements of EP-IP-440 EMERGENCY EXPOSURE CONTROLS?
A. 1.0 Rem B. 4.5 Rem C. 5.0 Rem D. 10.0 Rem Proposed Answer: C Explanation (Optional):
A. Incorrect: Previous exposures do not matter once the emergency is declared. Plausible in that this response is the 5 Rem minus the life time exposure.
B. Incorrect: Previous exposures do not matter once the emergency is declared. Plausible in that this response is the 5 Rem minus the exposure to date.
C. Correct: From EP-IP440: From the time an emergency is declared, ERO personnel are considered emergency workers. Emergency workers are allowed to receive the PNPS 2014 NRC Written 022014.doc
following exposure over the course of the emergency, exclusive of previous exposure and without special authorization: 5 Rem TEDE (whole body)
D. Incorrect: 10 Rem is the limit for operating equipment (the bridge) but is one of the higher exposure limits of EP-IP-440 that must be authorized by the ED/ EPM Technical Reference(s): EP-IP-440 Page 7 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank# X Modified Bank#
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.3.15 Importance Rating 2.9 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
KIA Justification: The C19 panels referred to in the stem contain fixed radiation monitors that will go into alarm on high particulate, iodine, gaseous activity or background radiation levels.
Any monitor reaching an alarm setpoint will result in the Trouble Alarm mentioned in the stem.
Proposed Question: RO Question # 71 The plant is at rated conditions. SBGT is running and the torus is being vented via the 2 inch torus vents, A0-5041 A and B, in order to adjust drywall to torus differential pressure.
While venting, annunciator C19 AlB Trouble, C904LC-B3 is received.
Which one of the following is correct?
A. Direct Radwaste to pump the drywall sumps. Venting may continue while drywall leakage rates are being verified to be within limits.
B. Direct Chemistry to verify the alarm locally. Venting may continue while the alarm is being verified.
C. Venting may continue provided the SBGT Disch Rad Hi alarm is not received.
D. Venting must be immediately secured.
Proposed Answer: D Explanation (Optional):
A. Incorrect: Venting must be immediately secured. Plausible in that venting may recommence if containment leakage rates are within Tech Spec limits.
B. Incorrect: Venting must be immediately secured. Plausible in that there is no direct readout of the C-19 Rad monitors in the control room.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: Venting must be immediately secured. Plausible in that venting is being performed using SBGT.
D. Correct: lAW PNPS Attachment 3, venting must be immediately secured.
Technical Reference(s): PNPS 2.2.70, Attachment 13, (Attach if not previously provided) pages 4 and 5 ARP C904LC-B3 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-08-02, TO VENT THE (As available)
PRIMARY CONTAINMENT DURING STATION OPS TO MAINTAIN PRIMARY CONTAINMENT PRESSURE WITHIN THE SPECIFIED PRESSURE REGULATION BAND.
Question Source: Bank # PNPS 6280 Made several changes but the question and answer remain the same.
Modified Bank #
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.4.18 Importance Rating 3.3 Knowledge of the specific bases for EOPs.
Proposed Question: RO Question # 72 Regarding EOP-03, Primary Containment Control:
Which one of the following is the bases for the requirement to initiate drywell sprays when torus bottom pressure exceeds 16 psig?
This ensures that. ..
A. The differential pressure rating of the containment vents is not exceeded.
B. Sufficient non condensable gases remain in the drywell to prevent chugging.
C. The negative design pressure of the containment will not be exceeded when sprays are initiated.
D. The reactor building to torus vacuum breakers do not lift resulting in oxygen being added to the primary containment.
Proposed Answer: B Explanation (Optional):
A. Incorrect: This is a function of the Primary Containment Pressure Limit B. Correct: Drywell sprays are initiated when torus bottom pressure exceeds 16 psig, the Suppression Chamber Spray Initiation Pressure (SCSIP), to preclude chugging - the cyclic condensation of steam at the downcomer openings.
If torus bottom pressure is greater than 16 psig, the possibility exists for "chugging" to occur. Chugging is the cyclic condensation of steam at the down comer openings of the drywell vents. This cyclic condensation causes stresses at the Downcomers which can lead to fatigue failure of the Down comers. This could cause any steam release in the drywell to bypass the torus and directly pressurize the torus airspace .
Chugging is a concern only if >99% of the non-condensibles (N2) in the drywell are in PNPS 2014 NRC Written 022014.doc
the torus. The torus bottom pressure of 16 psig is based on the lowest pressure that the torus would experience if 95% of the drywell non-condensibles were in the torus. When drywell sprays are initiated, the resulting pressure reduction opens the torus-to-drywell vacuum breakers, drawing non-condensibles from the torus back into the drywell.
C. Incorrect: This is a function of the Drywell Spray Initiation Limit.
D. Incorrect: This is the bases for securing drywell sprays before containment pressure lowers to 0 psig.
Technical Reference(s): 0-R0-03-04-05, Pages 40 and 41 (Attach if not previously provided)
Proposed References to be provided to applicants during examination:
Learning Objective: 0-R0-03-04-02, E0-26 (As available)
Question Source: Bank# WTSI # 11647 Modified Bank#
New Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9, 10 55.43 Shielding, isolation, and containment design features, including access limitations.
Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.4.29 Importance Rating 3.1 Knowledge of the Emergency Plan Proposed Question: RO Question# 73 You are a licensed RO on days off when the Emergency Plan is activated.
You have been contacted by the Emergency Plant Operations Supervisor (EPOS) and directed to support the Emergency Response Organization.
If the EPOS directs you to report and man the ENS, you will report to the_ (1) _ _.
If the EPOS directs you to report and man the PDP, you will report to the _ (2) _ _.
A. ( 1) EOF (2) Control Room B. (1) EOF (2) TSC C. ( 1) Control Room (2) Control Room D. ( 1) Control Room (2) TSC Proposed Answer: C Explanation (Optional):
A. Incorrect: The off-shift operator is directed to man the ENS in the control room.
Plausible in that there is an ENS line in the EOF but it is manned by non-operations personnel.
B. Incorrect: The off-shift operator is directed to man the ENS in the control room.
Additionally the off-shift operator is directed to man the PDP in the control room.
Plausible in that there is a PDP in the TSC but it is manned by non-operations personnel.
PNPS 2014 NRC Written 022014.doc
C. Correct: lAW EPIP 210, the EPOS is required to call-out one operator to man both the ENS and the PDP in the control room.
D. Incorrect: The off-shift operator is directed to man the PDP in the control room.
Technical Reference(s): EPIP-210, Control Room (Attach if not previously provided)
Augmentation, pages 4 and 5.
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G 2.1.18 Importance Rating 3.6 Ability to make accurate, clear and concise logs, records, status boards, and reports.
Proposed Question: RO Question # 74 A surveillance procedure is in progress that directs that an electrical lead be lifted during the performance of the surveillance.
Which of the requirements below must be met in order for the lifting of the lead NOT to be logged in the Lifted Lead and Jumper Log?
Requirement 1: The procedure contains signoffs for the initial lifting and subsequent re-landing of the lead Requirement 2: The lifted lead is re-landed and returned to its normal configuration prior to the end of the Shift Manager's shift.
Requirement 3: The surveillance results in an Active LCO and the LCO is tracked via the ESOMs software.
A. Requirement 2 only B. Requirement 3 only C. Requirements 1 and 2 D. Requirements 1 and 3 Proposed Answer: C Explanation (Optional):
A. Incorrect: The procedure must also contain signoffs for removal and relanding of the lead.
B. Incorrect: This is not a requirement. Plausible in that this is a configuration control process.
C. Correct: An electrical lifted lead, jumper, or gagging device need not be entered in the LLJ Log if the activity is performed in accordance with an approved Station Procedure PNPS 2014 NRC Written 022014.doc
provided that:
- The placing of the jumper, gagging device, lifting of the lead and the removal of the jumper, gagging device, landing of the lead are specific signoff steps in the approved governing Procedure.
AND
- The jumper, gagging device, or lifted lead will be returned to its normal configuration prior to the end of the current SM working shift.
D. Incorrect: Requirement 3 is not required. Requirement 2 is.
Technical Reference(s): PNPS 1.5.9.1, page 6 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01, 18c (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.2.15 Importance Rating 3.9 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.
Proposed Question: RO Question # 75 With the plant at 50% power, a closed motor operated valve located in the drywell must be tagged in the closed position.
Which one of the following is correct regarding the verification that the valve is in the closed position and tagged closed?
A. A non-independent verification is performed by observing the indicated valve position in the control room by two operators. A non-independent verification that the correct breaker has been identified is performed before the breaker is opened.
B. A non-independent verification is performed by observing the indicated valve position in the control room by two operators. The breaker for the valve is then opened and independently verified open.
C. The valve is verified closed by observing the indicated valve position in the control room. The position indication is then independently verified. The breaker for the valve is then opened and independently verified open.
D. The valve is verified closed by observing the indicated valve position in the control room. The position indication is then independently verified. A non-independent verification that the correct breaker has been identified is performed before the breaker is opened.
Proposed Answer: D Explanation (Optional):
A. Incorrect: An independent verification is required using the indirect method by observing the valve's position indication.
B. Incorrect: An independent verification is required using the indirect method by observing the valve's position indication. Additionally, two operators must use concurrent verification that the correct breaker has been identified before opening the breaker.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: Two operators must use concurrent verification that the correct breaker has been identified before opening the breaker.
D. Correct: Because the valve is in the drywall, the position of the valves is to be independently verified using the indirect method by observing the position indication in the control room. The opening of the correct breaker must be verified by two operators.
PNPS 1.3.34 requires non-independent verification in circumstances where the action (opening a breaker) is irreversible.
Technical Reference(s): 1.3.34, Pages 42 & 43 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01 (As available)
Question Source: Bank # PNPS bank 3362 Reworded all distractors, replaced two Modified Bank #
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295023 AA2.04 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS :
Occurrence of fuel handling accident Proposed Question: SRO Question # 76 The plant is refueling. The bridge is over the core with a fuel bundle loaded on the grapple.
The grapple is in the full-up position.
Then, Refuel SRO reports a lowering pool level.
Additional indications are:
- Annunciator Spent Fuel Pool Level Lo (C903R-B2) has alarmed
- Annunciator Refuel Floor Rad Hi (C904LC-C7) has alarmed o ARM Spent Fuel Pool Area is determined to be the cause of the alarm
- No other Reactor Building ARMs are in alarm
- Standby Gas Treatment trains auto start The SRO on the refuel floor is required to direct __ (1) _ _.
The SRO in the control room is required to direct __ (2) _ .
A. (1) the bundle be lowered into the nearest Reactor Vessel location with a fully inserted control rod (2) CRHEAFS be started B. ( 1) the bundle be lowered into the nearest Reactor Vessel location. The location need not contain a fully inserted control rod.
(2) CRHEAFS be started C. ( 1) the bundle lowered into the nearest Reactor Vessel location with a fully inserted control rod (2) the reactor building be evacuated D. ( 1) the bundle be lowered into the nearest Reactor Vessel location. The location need not contain a fully inserted control rod.
(2) the reactor building be evacuated PNPS 2014 NRC Written 022014.doc
Proposed Answer: A Explanation (Optional):
A. Correct: Both PNPS 5.4.3, Refuel Floor High Radiation and PNPS 2.4.31, Reactor Basin and/or Spent Fuel Pool Drain-Down are required to be entered. Both of these procedures require that IF a fuel bundle is grappled onto the Refuel Bridge mast, THEN IMMEDIATELY LOWER the fuel bundle into the nearest Reactor Vessel location with a fully inserted control rod OR the nearest Spent Fuel Pool location. Since the bridge is over the core the bundle should be lowered back into the vessel.
PNPS 5.4.3, subsequent actions, requires that the Control Room High Efficiency Air Filtration System (CRHEAFS) be started.
B. Incorrect: The vessel location must contain a control rod. Plausible if the candidate believes that there is always sufficient S/D margin available to tolerate one bundle being lowered into a cell without a control rod and that lowering the bundle takes precedence.
C. Incorrect: Both procedures require that the refuel floor be evacuated. Plausible in that PNPS 5.4.3, subsequent actions, does direct that the reactor building be evacuated under certain conditions. These are:
- Both SBGT trains failed
- Any Reactor Building ARM in alarm other than those on the refuel floor.
Neither of these conditions exists so the reactor building need not be evacuated. Since local actions may need to be taken to mitigate the lowering pool level, the reactor building should only be evacuated if necessary.
D. Incorrect: The fuel bundle is to be lowered into in the nearest Reactor Vessel location with a fully inserted control rod. Additionally, there is no requirement do evacuate the reactor building. Since local actions may need to be taken to mitigate the lowering pool level, the reactor building should only be evacuated if necessary.
Technical Reference(s): PNPS 5.4.3, sections 3.0 and 4.0 (Attach if not previously provided)
PNPS 2.4.31, section 3.0 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-04-03, EOs 9a and 1Oa (As available}
Question Source: Bank#
PNPS 2014 NRC Written 022014.doc
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 7 Fuel handling facilities and procedures.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295030 EA2.04 Importance Rating 3.7 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Drywell/ suppression chamber differential pressure: Mark-1&11 Proposed Question: SRO Question # 77 The plant is at rated conditions with the following Primary Containment parameters.
- Torus water level is -6 inches
- Drywell to Torus differential pressure is 1.17 psid If Torus water level CONTINUES to LOWER which one of the following is required by Tech Specs?
A. Initiate an orderly shutdown and be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No other Tech Spec action is required.
B. Restore the drywell to torus differential pressure within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. No other Tech Spec action is required.
C. Initiate an orderly power reduction to < 15% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. Initiate an orderly shutdown and be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND restore the drywell to torus differential pressure within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Proposed Answer: D Explanation (Optional):
A. Incorrect: A differential pressure of 1.17 is also at the lower Tech Spec limit as specified by TS 3.7.A.8.a. lftorus water level lowered further the drywell differential pressure would also lower. TS 3.7.A.8.c then requires that the differential pressure must be restored within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If not then the reactor shall be less than 15% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. Incorrect: A plant shutdown must also be initiated. This would be the correct answer if the candidate only applied the TS action for the out of spec differential pressure.
PNPS 2014 NRC Written 022014.doc
C. Incorrect: This would be true if the candidate determined that the drywall to torus differential pressure could not be restored within B hours.
D. Correct: lAW TS 3. 7.A.1, the minimum Torus water level is B4,000 fe. This corresponds to a torus level of -6 inches. This is also addressed in TS 3.7.A.1.m. lftorus water level continued to lower both of these limits would be exceeded. TS 3. 7.A.5 then requires that an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
A differential pressure of 1.17 is also at the lower Tech Spec limit as specified by TS
- 3. 7 .A.B.a. If torus water level lowered further the drywall differential pressure would also lower. TS 3. 7.A.8.c then requires that the differential pressure must be restored within the next B hours. If not then the reactor shall be less than 15% within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Technical Reference(s): TS 3. 7.A.1 (Attach if not previously provided) 3.7.A.B.a TS Bases page B3/4. 7-B Proposed References to be provided to applicants during TS 3. 7 .A.1 through A.B.
examination: Blackout sections 3.7.A.1.c through j, and 3. 7.A.2.
Learning Objective: (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295031 EA2.02 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL : Reactor power Proposed Question: SRO Question # 78 A failure to scram has occurred and the following plant conditions exist:
- Reactor power is 20% and lowering
- RPV water level is +5 inches and lowering
- Both Recirc Pumps have been tripped
- Drywell pressure is 3.2 psig
- Torus bulk water temperature is 112 oF and rising
- Injection has been terminated and RPV level is being intentionally lowered lAW EOP-02 RPV Control Failure to Scram, which ONE of the following set of conditions would allow recommencing injection?
Power lowering to ...... .
A. 4% with RPV level lowering to -15 inches B. 6% with RPV level lowering to -125 inches, actual C. 8% with RPV level lowering to -100 inches, actual D. 10% with RPV level lowering to -30 inches Proposed Answer: 8 Explanation (Optional):
A. Incorrect: Once lowering of level is commenced injection is not recommenced until level is below -25 inches regardless of how much power has dropped. This is to prevent I mitigate power oscillations. Plausible in that power has dropped the SliT curve ( < 10%)
PNPS 2014 NRC Written 022014.doc
B. Correct: Even though reactor power is still above 3%, injection is recommenced when RPV level drops reaches -125 inches in order to maintain adequate core cooling.
C. Incorrect: Level must continue to be lowered because power is still above 3%.
Plausible in that power has dropped below 10% which is the power level associated with exceeding the Boron Injection Initiation Temperature (BIIT) curve and exceeding the BIIT curve is a criterion for intentionally lowering level.
D. Incorrect: Level must continue to be lowered because power is still above 3%.
Plausible in that level is below -25 inches which is that RPV level associated with preventing or mitigating power oscillations. However the conditions presented require level to be lowered further to protect the containment.
Technical Reference(s): EOP-02, steps L-14 thru L-21 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: EOP-02, "Q" leg only Learning Objective: 0-R0-03-04-04, 23g (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295018 2.1.19 Importance Rating 3.8 Conduct of Operations: Ability to use plant computers to evaluate system or component status.
(Partial or Total Loss of CCW)
Proposed Question: SRO Question# 79 The plant is at rated power when a large mat of seaweed begins to clog the intake structure.
Current conditions in the screenhouse are as shown below on EPIC screen 504.
Which one of the following is required?
A. Insert a manual scram, enter PNPS 2.1.6, Reactor Scram and secure both seawater pumps.
B. Enter PNPS 2.1.14, Station Power Changes, reduce power as required and then secure the "B" Seawater pump.
PNPS 2014 NRC Written 022014.doc
C. Immediately secure the "B" seawater pump. Reduce power as required to maintain condenser vacuum lAW PNPS 2.4.36, Decreasing Condenser Vacuum. Securing SSW pumps is not required.
D. Immediately secure the "B" seawater pump. Reduce power as required to maintain condenser vacuum lAW PNPS 2.4.36, Decreasing Condenser Vacuum and place the "A", "B" and "C" SSW pumps in Pull-to-Lock.
Proposed Answer: C Explanation (Optional):
A. Incorrect: A manual scram is not yet required as the "A" seawater pump has not yet reached a condition requiring that it be tripped or secured. Plausible in that it is approaching that level.
B. Incorrect: The pump must be immediately tripped to preserve suction to the SSW pumps. Plausible in that there is a step to reduce power and then secure the pump.
However level is too low to allow time to reduce power prior to tripping the pump.
C. Correct: lAW PNPS 2.4.154, if any of the following Screenweii/Seawater Pump bay levels are observed, THEN IMMEDIATELY STOP the affected Seawater Pump in accordance with Step 4.0[4]: Screenweii/Seawater Pump bay level has dropped to below the -10 foot elevation. Tripping the SSW pumps is not required as they will retain suction down to -13 feet, 9 inches and level should quickly recover once the seawater pump is secured.
D. Incorrect: Tripping the SSW pumps is not required as they will retain suction down to
-13 feet, 9 inches and level should quickly recover once the seawater pump is secured.
Technical Reference(s): PNPS 2.4.154, section 4.0 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-03-23(07), TO RESPOND (As available)
TO INTAKE STRUCTURE FOULING FROM THE CONTROL ROOM.
Question Source: Bank #
Modified Bank #
New X PNPS 2014 NRC Written 022014.doc
Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295025 2.2.44 Importance Rating 4.4 Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives effect plant and system conditions. (High Reactor Pressure)
Proposed Question: SRO Question # 80 Which of the following sets of parameters would result in exceeding the Primary Containment Pressure Limit during an emergency depressurization?
HEAT CAPACITY TEMPERATURE LIMIT 6 PRESSURE SUPPRESSION PRESSURE SPDS033 200 Torus Water levels ..;.127 in.
SPDS 038 260 G: 240 t_,
e i 220 1----'
~
~ 200 r------t-----'
I-
.e
~ 180 1----*-4-******-************f*********~ ..**-+**....""~-
~ 160 !---""'f....------l----1--*---+--~
- 100 120 !40 160 :HID 20C 200 400 600 000 1000 90 Torus Water Level (in.) 175 RPV Pressure {psig}
A. RPV pressure is 700 psig Torus water level is 89" B. Torus water level is 140" Torus bottom pressure is 31 psig C. Torus water temperature is 181 oF RPV pressure is 700 psig Torus water level is 95" PNPS 2014 NRC Written 022014.doc
D. Torus water temperature is 160°F RPV pressure is 1050 psig Torus water level is 125" Proposed Answer: C Explanation (Optional):
A. Incorrect: The PCPL will not be exceeded if an emergency depressurization (ED) is performed with these parameters. Plausible in that the Torus Downcomers are uncovered at this level and EOP-03 requires an ED at this value. However the SRV tail pipes are still covered at this value.
B. Incorrect: The PCPL will not be exceeded if an emergency depressurization (ED) is performed with these parameters. Plausible in that the PSP is being exceeded at these values.
C. Correct: The 3 parameters provided result in exceeding the HCTL.. The Heat Capacity Temperature Limit (HCTL) is the highest torus water temperature from which emergency RPV depressurization will not raise:
- Torus water temperature above the maximum temperature capability of the torus and equipment within the torus which may be required to operate when the RPV is pressurized, or
- Torus bottom pressure above the Primary Containment Pressure Limit (PCPL),
while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
For PNPS, the second criterion defines the HCTL.
D. Incorrect: The HCTL is not being exceeded. Plausible if the candidate uses the lower limit for torus water~ 91 inches but < 100 inches, in which case the candidate would think that the limit was being exceeded.
Technical Reference(s): 0-R0-03-04-05, Page 29 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-05, E09 (As available)
PNPS 2014 NRC Written 022014.doc
Question Source: Bank # WTSI Bank 3717 Modified Bank#
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295019 2.1.32 Importance Rating 4.0 Conduct of Operations: Ability to explain and apply all system limits and precautions. (Partial or Total Loss of lnst. Air)
Proposed Question: SRO Question # 81 The reactor is at rated conditions, with the instrument air system aligned for normal full power operation.
Then, a loss of instrument air occurs. Procedure 5.~1.8, Loss of Instrument Air is entered.
Which one of the following is correct regarding the 5.3.8 procedure limit below, the bases for that limit and the required action?
If Instrument Air Main Header pressure is less than 65 psig, then enter procedure ...
A. 2.1.6 Reactor Scram and manually scram the reactor. The ADS accumulators will not have sufficient volume for the required number of actuations at this pressure.
B. 2.1.6 Reactor Scram and manually scram the reactor. Control rods may begin drifting into the core at this pressure.
C. 2.4.49, Feedwater Malfunctions and take manual control of the Feed Reg Valves in the condenser bay. The Feed Reg Valves will lock up at this pressure.
D. 2.4.49, Feedwater Malfunctions and reduce Reactor Power to stabilize reactor water level lAW Section 7.11 of PNPS 2.1.14, "Station Power Changes". The Condensate Minimum Flow Valve will fail open at this pressure reducing available condensate flow.
Proposed Answer: B Explanation (Optional):
A. Incorrect: PNPS 5.3.8 requires a scram based on rods drifting in. Plausible in that the procedure addresses the ADS accumulators. However the pressure is where the ADS accumulators have insufficient volume is 91 psig. The required action is to close A0-4356, Essential Instrument Air Block Valve if instrument air is aligned to the drywell and pressure falls below this value. At rated conditions this would not be the case.
PNPS 2014 NRC Written 022014.doc
B. Correct: lAW the Subsequent Action of procedure 5.3.8, the reactor is to be manually scrammed at this air pressure. This is because the scram air header pressure is lowering and the scram outlet valves may begin to open causing control rods to drift into the core.
C. Incorrect: The procedure action at this pressure is to scram the reactor. Plausible in that the FRVs will lockup on lowering air pressure. However this will occur at 40 psig.
D. Incorrect: The procedure action at this pressure is to scram the reactor. Plausible in that the Condensate Min Flow valve will be affected as pressure lowers. However the valve fails closed. If it were to fail open the impact would be as described.
Technical Reference(s): 5.3.8, subsequent actions (Attach if not previously provided)
ARP C905R-F1 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-02-28(01 ), Terminal (As available)
Objective, RESPOND TO A LOSS OF INSTRUMENT AIR.
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 295004 2.2.25 Importance Rating 4.2 Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits (Partial or Total Loss of DC Pwr)
Proposed Question: SRO Question # 82 The plant is at rated conditions with 125 VDC Battery Backup Charger D14 inoperable. All other equipment is operable.
Then, the "B" 125 VDC Battery Charger fails. "B" Battery terminal voltage is later reported as having lowered to 11 0 VDC.
Assuming that conditions do not change the plant must be in Cold Shutdown i n _ (1) _ _
hours.
This is because the battery is not capable supplying the required loads over the _ (2) _
hour service load period.
A. (1) 24 (2) 12 B. (1) 72 (2) 12 C. (1) 24 (2) 8 D. (1) 72 (2) 8 Proposed Answer: C Explanation (Optional):
A. Incorrect: The service load period is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
B. Incorrect: The plant must be cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Plausible because Tech Spec 3.9.B.5 allows 3 days of continued operation if TS 3.5.F is met. However 3.5.F is not met.
PNPS 2014 NRC Written 022014.doc
Additionally, the service load period is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C. Correct: A battery terminal voltage of equal to or greater than 119.0 volts for the 125 volt system under normal station load is capable of supporting bounding accident shutdown loads over the 8-hour service load period. The battery terminal voltage will decrease further with higher load levels but will still remain capable of performing its design function, based on the Pilgrim DC load flow studies results. Therefore the battery is inoperable.
lAW Tech Spec 3.9.B.5 from and after the date that one of the 125 or 250 volt battery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding three days within electrical safety considerations, provided repair work Is initiated in the most expeditious manner to return the failed component to an operable state, and Specification 3.5.F is satisfied. However TS 3.5.F.1 is not satisfied because the "B" EDG and the "B" side low pressure core and containment systems are inoperable. TS 3.5.F.1 requires that the plant be in cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if this LCO is not met.
D. Incorrect: The plant must be cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Plausible because Tech Spec 3.9.B.5 allows 3 days of continued operation if TS 3.5.F is met. However 3.5.F is not met.
Technical Reference(s): Tech Spec 3.9.B.5 and (Attach if not previously provided) associated bases.
TS 3.5.F PNPS 5.3.12, Discussion item [1]
and [2]
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-01-02, EO 15 and 16 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 PNPS 2014 NRC Written 022014.doc
55.43 2 Facility operating limitations in the technical specifications and their bases.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295020 AA2.02 Importance Rating 3.4 Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION: Drywell/containment temperature Proposed Question: SRO Question# 83 The plant is at rated power when the following sequence of events occurs.
Time:
T=O RBCCW valve M0-4009A, RBCCW Loop B Nonessential Loop Inlet Valve is inadvertently closed T=+ 5min The reactor is manually scrammed when M0-4009A cannot be reopened T=+ 8 min Drywell pressure exceeds 2.2 psig T=+15 min Drywell temperature exceeds 250 oF T=+ 25 min Ability to reopen M0-4009A has been restored but the valve has not yet been reopened.
Given the above conditions which one of the following is the correct regarding:
(1) Whether RBCCW can or cannot be restored to the drywell AND (2) Any required actions?
A. ( 1) Cannot (2) The only method to control drywell temperature is lAW EOP-03, Primary Containment Control.
B. (1) Can (2) Restore flow by re-opening M0-4009A lAW Attachment 7, Opening RBCCW Valve M0-4009A of PNPS 2.4.42, Loss of RBCCW.
No other action is required.
C. (1) Can (2) Restore flow by performing Section 4.2 Recovery of RBCCW Loop "B" with an Elevated Drywell Temperature of PNPS 2.4.42, Loss of RBCCW.
PNPS 2014 NRC Written 022014.doc
D. (1) Cannot (2) Defeat the Group 2 isolation signal and commence a venting and purging of the drywell with air lAW PNPS 2.4.44, Loss of Drywell Cooling.
Proposed Answer: C Explanation (Optional):
A. Incorrect. RBCCW flow can be restored. Plausible in that there are two parts to PNPS 2.4.42, Loss of RBCCW, Section 4.2 RECOVERY OF RBCCW LOOP "B" WITH AN ELEVATED DRYWELL TEMPERATURE. Both sections deal with drywell temperatures over 250 degrees. However the 1st section is utilized if the drywell temperature is also affiliated with a LOCA. In this case the RBCCW loop is recovered but flow is not restored to the drywell. If flow could not be restored the required actions would be as described in EOP-03.
B. Incorrect: With RBCCW flow having been loss for > 6 minutes and drywell temperature being > 250 degrees, there is a potential for a water hammer event when flow is restored. The process of 2.4.42, Loss of RBCCW, Section 4.2 must also be followed.
C. Correct: With RBCCW flow having been loss for > 6 minutes and drywell temperature being > 250 degrees, there is a potential for a water hammer event when flow is restored. The process of 2.4.42, Loss of RBCCW, Section 4.2 must be followed. M0-4009A must then be re-opened using Attachment 7 of this procedure.
D. Incorrect: RBCCW flow can be restored if the process described in section 4.2 of 2.4.42 is followed. Also PNPS 2.4.44 does address venting the containment but there is no authorization to defeat the Group II isolation in that procedure to facilitate doing so.
Technical Reference(s): PNPS 2.4.42, section 4.2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-02-06, E011 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used PNPS 2014 NRC Written 022014.doc
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 500000 2.4.8 Importance Rating 4.5 Emergency Procedures I Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOP's (High CTMT Hydrogen Cone)
Proposed Question: SRO Question # 84 EOP-03, Primary Containment Control is being executed. Plant conditions are as follows:
- All control rods inserted.
- Reactor level: -70 inches, actual, and slowly rising via CRD injection.
- RPV Pressure: 375 psig lowering
- RPV level was below -150 inches, actual, for a period of time before beginning to recover
- H2 injection was isolated following the scram
- Neither H2 0 2 analyzer can be placed in service lAW PNPS 5.7.3.2, Drywell and Torus Atmospheric Sampling under tF H2 or 02 r11011J!oring system is or : THEfl o;ampllo d!yW<!l! and torus for H2 becomes unavaaabfe : andOz Emergency Conditions and EOP-03, procedure 5 .7. 3.2 which one of the following is correct IF a reactor scram has been initiated : THEN verify Hz injecti<m is secuM i and isolated (1) regarding EOP-03 combustible gas control, Start and operate H2 and Oz an>itzerS,
~ ... 2.2.133 AND (2) any required EOP-03 action?
1%
Redu"" primary coni.\mmMt H2
"'"'e.,lmtion, proced""' 2.2.70 A. ( 1) Hydrogen concentration can be assumed to be < 1%
(2) PASS sampling is not required.
B. (1) Hydrogen concentration cannot be determined.
(2) Direct a PASS sample be obtained before taking additional action.
PNPS 2014 NRC Written 022014.doc
C. ( 1) It must be assumed that Hydrogen concentration is > 1%
(2) Commence venting the containment lAW procedure 2.2. 70, Primary Containment Control. Exceeding release limits is authorized.
D. (1) It must be assumed that Hydrogen concentration is> 1%
(2) Commence venting the containment lAW procedure 2.2. 70, Primary Containment Control. Exceeding release limits is not authorized.
Proposed Answer: D Explanation (Optional):
A. Incorrect: Plausible in that RPV level is currently -70 inches and recovering. PNPS 5.7.3.2 also directs that if all the following conditions are met then PASS sampling is NOT required:
- The Reactor is shutdown and will remain shutdown under all conditions;
- Initial oxygen concentration in the Drywell and Torus was 3% or less;
- The Torus-to-Reactor Building vacuum breakers have not opened since oxygen
- concentration was established at 3% or less; Since there is no evidence that any of these conditions were not met, PASS sampling is not required.
B. Incorrect: It must be assumed that [H2] is> 1% since RPV level dropped below TAF see below. Therefore venting is required.
C. Incorrect: Exceeding release rates is not authorized as the Release Icon is not associated with this step.
D. Correct: lAW the discussion section of PNPS 5.7.3.2, failure or unavailability of hydrogen or oxygen monitoring system does not necessarily mean that hydrogen and oxygen concentrations cannot be determined. Rather, Operator judgment is required when making that decision after examining related plant conditions. If the Operator has reasonable assurance that a transient has not occurred which has uncovered the core (i.e., water level has not dropped below TAF), then it is likely that significant hydrogen has not been produced. On the other hand, if an event has occurred and water level cannot be determined or did, in fact, drop below TAF, then the Operator has reason to be concerned that significant amounts of hydrogen may have been generated unless alternate methods of determination indicate otherwise. Alternate methods include sampling, which may involve a long time delay, or deduction based on actual plant status.
This is further reinforced in Discussion Item [2](c) (2) which says that Primary Containment hydrogen cannot be determined to be below 1% when RPV water level has dropped below top of active fuel.
Since it cannot be assumed that [H2] is < 1% venting is required per steps G-3 and G-4 PNPS 2014 NRC Written 022014.doc
of EOP-03. However exceeding release rates is not authorized as the Release Icon is not associated with this step.
Technical Reference(s): PNPS 5.7.3.2 page 5 discussion (Attach if not previously provided) section and item [2]
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-05, EO 14 a. b (As available)
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 295007 AA2.01 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor pressure Proposed Question: SRO Question # 85 The plant was at rated conditions when a high drywell pressure condition resulted in a reactor scram. Current plant conditions are as follows:
- Torus Bottom pressure: 20 psig, rising slowly
- Drywell temperature: 250 degrees, rising slowly
- Rx Water level: 25 inches, steady
- Rx pressure: 935 steady
- All control rods are inserted
- The MSIVs are open
- Drywell sprays are about to be placed in service lAW with plant procedures which one of the following RPV pressure control strategies should be implemented?
A. lAW EOP-01, RPV Control, commence a cooldown by partially opening one main turbine bypass valve and slowly but steadily reduce pressure not to exceed a cooldown rate of 100 degrees/hour.
B. lAW PNPS 5.3.35.2, Operations Emergency and Transient Response Strategies, fully open one main turbine bypass valve and rapidly lower pressure. Stabilize pressure between 450 to 550 psig.
C. lAW EOP-01, RPV Control, anticipate an Emergency Depressurization and open all bypass valves.
D. lAW EOP-17, Emergency RPV Depressurization, open all SRVs and Emergency Depressurize.
Proposed Answer: B Explanation (Optional):
PNPS 2014 NRC Written 022014.doc
A. Incorrect: The strategy is to aggressively reduce RPV pressure using available Table M systems when a primary system leak is threatening the containment.
B. Correct: lAW PNPS 5.3.35.2, Attachment 2, If MHC is available, fully open one main turbine bypass valve using the bypass valve opening jack while maintaining water level within -20 to +45" band. Stabilize pressure 450 to 550 psig to avoid exceeding a 100°F/hour cooldown rate.
C. Incorrect: Although containment parameters are high plant conditions do not yet currently warrant that the ED be anticipated as drywell sprays have not yet been placed in service.
D. Incorrect: Although containment parameters are high plant conditions do not yet currently require an Emergency Depress (ED). The criteria for an ED is that the parameter cannot be maintained below its limit, and without first attempting drywell sprays this determination cannot be made.
Technical Reference(s): PNPS 5.3.35.2, Attachment 2, (Attach if not previously provided)
Sequence 1 and 2 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-02, EO 29 and 30 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 209001 A2.05 Importance Rating 3.6 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Core spray line break Proposed Question: SRO Question # 86 Given the following:
- The plant is at rated conditions
- HPCI is out of service for maintenance and the plant is on day 2 of a 14 day LCO per Tech Spec 3.5.C
- The non-licensed operator reports that Core Spray "A" Line break differential pressure, dPIS-1459A, is indicating +5.0 psid Assuming that conditions do not change, which one of the following (1) is correct regarding the status of Core Spray "A" AND (2) is correct per Technical Specifications?
A. (1) The "A" Core Spray injection line has failed inside the shroud.
(2) Plant operation can continue for 7 days. The plant shall then be in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B. (1) The "A" Core Spray injection line has failed between the vessel wall and the shroud.
(1) Plant operation can continue for 7 days. The plant shall then be in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. (1) The "A" Core Spray injection line has failed inside the shroud.
(2) An orderly plant shutdown is required and the plant shall be in cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. (1) The "A" Core Spray injection line has failed between the vessel wall and the shroud.
(2) An orderly plant shutdown is required and the plant shall be in cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
PNPS 2014 NRC Written 022014.doc
Proposed Answer: D Explanation (Optional):
A. Incorrect: Each core spray system has an instrument line that is connected to the low pressure side of a differential pressure switch, located on instrument rack 2207. These switches (dPIS-1459A/B) provide an alarm when a core spray line breaks downstream of the injection line check valve. The high pressure side of the dPIS is connected to the standby liquid control injection "outer pipe, which detects the pressure in the bypass region above the core plate. If a CS line breaks inside the shroud, the dPIS low pressure side will detect reactor pressure inside the shroud as usual. The CS sparger fracture inside the shroud will not cause an alarm. Additionally, per TS 3.5.C, from and after the date that the HPCI system is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding 14 days unless such system is sooner made operable, providing that during such 14 days all active components of the ADS system, the RCIC system, the LPCI system and both core spray systems are operable. If these requirements cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B. Incorrect: With Core Spray "A" inop in conjunction with HPCI, the plant must be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C. Incorrect: The line has failed between the vessel wall and the shroud. See "A" above.
D. Correct: If the CS line breaks outside the core shroud, but inside the reactor vessel, the pressure on the low side is now the pressure outside the core shroud. There is an additional pressure drop {about 7.5 psi) across the steam separators and dryers. If we assume that normal operating sensed dP is -3.0 psid and the low side pressure is decreased by 7.5 psi, we obtain +4.5 psid. The alarm setpoint is -2.5 ~0.5 psid, and will therefore cause an alarm in the control room. Per TS 3.5.C, from and after the date that the HPCI system is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding 14 days unless such system is sooner made operable, providing that during such 14 days all active components of the ADS system, the RCIC system, the LPCI system and both core spray systems are operable. If these requirements cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Technical Reference(s): Core Spray reference text, page (Attach if not previously provided) 18 TS 3.5.C, PNPS 2014 NRC Written 022014.doc
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-02, EO 2g and 15 (As available)
Question Source: Bank# WTS 1190 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 203000 A2.12 Importance Rating 2.7 A2.12 -Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump runout Proposed Question: SRO Question# 87 EOP-26, RPV Flooding, Failure to Scram is being executed. The following conditions exist.
- All SRVs are open
- The Minimum Steam Cooling Pressure has been established
- RPV pressure is being maintained just above the required Minimum Steam Cooling Pressure via RHR injection
- RHR pumps "A" and "C" are injecting
- M0-1 00 1-28A, LPCI Injection Valve #1 , is throttled
- Total RHR injection flow is 6000 gpm
- Both Core Spray pumps are available but in standby
- No other sources of injection are available Then, RHR Pump "A" trips.
Which one of the following is correct?
Pressure will fall below the Minimum Steam Cooling Pressure (MSCP) and .....
A. can be re-established by throttling open M0-1 001-28A, LPCI Injection Valve #1.
B. cannot be re-established using RHR. Start Core Spray pumps as required to re-establish the MSCP.
C. cannot be re-established using RHR. Core Spray injection is prohibited. Exit EOP-26 and enter SAGs.
D. can be re-established by opening additional vent paths lAW EOP-26, Table P, Alternate RPV Depressurization Systems which will lower the required MSCP.
Proposed Answer: B PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: The single RHR pump cannot inject at 6000 gpm without going into runout.
Even if the candidate decides that exceeding the limit is necessary, the runout condition will result in very low discharge pressures preventing the re-establishment of the MSCP.
B. Correct: Prior to the pump trip each pump was injecting at 3000 gpm in order to maintain the required MSCP (240 psig w/4 SRVs open). One RHR pump cannot inject 6000 gpm without going into runout as the runout limit is 5600 gpm. Even if the candidate decides that exceeding the limit is necessary, the runout condition will result in very low discharge pressures preventing the re-establishment of the MSCP.
Step F-11 of EOP-26 directs the operator to establish the MSCP using outside the shroud systems and if necessary Table X systems. Although not preferred, Core Spray injection is required in order to avoid core damage.
C. Incorrect: Core Spay injection is allowed if required. SAGs are entered when core damage is indicated. Plausible in that Core Spray injection is avoided if at all possible during ATWS conditions as the water is not pre-heated and may "flush out" any boron that has been injected.
D. Incorrect: Injecting with Core Spray is the required action. Plausible in that alternate vent systems will be used if the MSCP cannot be established depending upon SRV status. However the other criteria for using alternate vent systems is if 4 SRVs cannot be opened.
Technical Reference(s): PNPS 2.2.19 Precaution# 2 (Attach if not previously provided)
EOP-26, step F-11, Tables Y and X
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-04-08, E013 (As available)
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PNPS 2014 NRC Written 022014.doc
10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215004 2.4.30 Importance Rating 4.1 Emergency Procedures I Plan; Knowledge of events related to system operation I status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator. (Source Range Monitor)
Proposed Question: SRO Question # 88 The plant is refueling with a core alteration in progress when the following sequence occurs:
Time: 00:00
- SRM counts begin to rise
- Sustained SRM Period alarms are received on two channels
- SRM counts continue to rise
- Manual Reactor scram inserted due to rising counts on multiple channels
- SRM counts lower to original values
- Refuel Floor Vent Exhaust monitors begin to rise Time: 00:20
- Refuel Floor Vent Exhaust Hi Alarms are received on all 4 channels .
- Main Stack Low Range Rad monitors RM-1705-18A & 8 begin to rise Time: 00:40
- Main Stack Low Range Rad monitors RM-1705-18A & B indicate 4E+5 cps and rising slowly Time 00:55
- Main Stack Low Range Rad monitors RM-1705-18A & 8 indicate 8E+5 CPS and are stable lAW EPIP-100 Emergency Classification And Notification, which one of the following describes ALL the reports to be made to the state and local agencies and the bases for those reports?
A. Initial Notification of an Alert based on exceeding an Alert EAL at 00:20 B. Initial Notification of an Unusual Event (UE) based on exceeding an UE EAL at 00:00 Follow-up Notification due to having a release in progress at 00:40 C. Initial Notification of an Unusual Event (UE) based on exceeding an UE EAL at 00:00 Initial Notification of an Alert based on exceeding an Alert EAL at 00:55 D. Initial Notification of an Unusual Event (UE) based on exceeding an UE EAL at 00:00 Initial Notification of an Alert based on exceeding an Alert EAL at 00:20 PNPS 2014 NRC Written 022014.doc
Follow-up Notification based on having a release in progress at 00:55 Proposed Answer: D Explanation (Optional):
A. Incorrect: A UE was exceeded at time 00:00 based on EAL CU5.1, Unplanned sustained positive period observed on nuclear instrumentation. Also a Follow-up Notification is required due to meeting the criteria of a Release in progress at time 00:55. Plausible if the Candidate does not recognize the UE or understand what a release in progress is.
B. Incorrect: An Alert was exceeded at time 00:20 when Refuel Floor Vent Exhaust Hi Alarms were received on all 4 channels. Also the criteria for declaring a release in progress was not met until 00:55. Time 00:40 is plausible for this since the level for the Alert Rad Release EAL was exceeded at this time. However that level must be present for 15 minutes.
C. Incorrect: The Alert was first exceeded at time 00:20 which would require an initial notification. Although an Alert EAL was also exceeded at time 00:55 due to Main Stack monitors being above 4E+5 for 15 minutes, an Initial Notification should have been sent out based on exceeding the Alert EAL at time 00:20.
D. Correct: An Unusual Event is required based on EAL CU5.1, unplanned sustained positive period observed on nuclear instrumentation. The initial SRM response indicates an unplanned criticality and a UE is required.
When Refuel Floor Vent Exhaust Hi Alarms were received on all 4 channels at time 00:20 an Alert is required based on EAL AA2.1, Damage to irradiated fuel OR loss of water level (uncovering irradiated fuel outside the RPV) that causes a valid high alarm on any of the following radiation monitors (Panel C91 O/C911 ):
- New Fuel Vault (RIS-1815-3D)
-Refuel Floor Shield Plug Area (RIS-1815-3E)
-Spent Fuel Pool Area (RIS-1815-3F)
- Refuel Floor Vent Exhaust (RIS-1705-BA-D)
Since the Refuel Floor Vent Exhaust monitors began to rise at the same time of the inadvertent criticality it must be assumed that damage to irradiated fuel has occurred.
At time 00:55 the criteria for declaring a release in progress was met. The definition of a release in progress is any release of radioactivity which meets any EAL of Subcategory A 1, Offsite Rad Conditions or Involves an actual or suspected Turbine Building or unmonitored release which is associated with the emergency event.
At time 00:55 Subcategory A 1 EAL AA 1.1, Any valid gaseous monitor reading >Table A-1 column "Alert" for 15 min was exceeded. At time 00:55, the Main Stack Rad monitors have been above the Alert Value of 4E+5CPS for 15 minutes.
PNPS 2014 NRC Written 022014.doc
Since an upgrade to a higher EAL is not warranted the information regarding the Release in Status is transmitted via a Follow-Up notification.
Technical Reference(s): Cold EAL Chart (Attach if not previously provided)
EPIP-100, definition [11] on page 7
Proposed References to be provided to applicants during examination: EAL Chart Learning Objective: 0-R0-07-02-01, EO 2 (As available)
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 215003 2.2.38 Importance Rating 4.5 Equipment Control: Knowledge of conditions and limitations in the facility license. (IRM)
Proposed Question: SRO Question # 89 The plant is refueling with the following plant conditions:
- All control rods are inserted
- Fuel is being loaded in the vessel Then, I&C reports that the IRM High Rod Block trip for aiiiRMs is set at 110/1251hs of scale.
Which one of the following is correct?
Core alterations ...
A. can continue but one IRM must be placed in the tripped condition within one hour.
B. must be immediately suspended until one IRM is placed in the trip condition.
C. can continue and a Tracking LCO must be generated.
D. must be immediately suspended.
Proposed Answer: C Explanation (Optional):
A. Incorrect: The IRM function is not required. Plausible in that this would be the required Tech Spec action if they were required.
B. Incorrect: There is no requirement to suspend core alterations until one IRM is placed in trip.
C. Correct: Per Table 3.2.C-2 of the FSAR, the IRM Upscale Rod Block setpoint must be <
108/125 of full scale. Therefore the IRM rod block function is inoperable for all IRMs.
However, the IRM function is not required in refuel if the head is removed and all rods are inserted per Note 6 of FSAR Table 3.2.C-1. Since they are not required, a Tracking PNPS 2014 NRC Written 022014.doc
LCO is required.
D. Incorrect: There is no requirement to suspend core alterations. Plausible in that the IRM function is required in refuel under certain conditions.
Technical Reference(s): FSAR Table 3.2.C-1, and (Attach if not previously provided) associated notes PNPS 1.3.34.2, Definition [12]
FSAR Table 3.2.C-2 Proposed References to be provided to applicants during examination: FSAR Table 3.2.C-1, and associated notes Learning Objective: 0-R0-02-07-02, EO 16 (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 7 Fuel handling facilities and procedures.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 264000 2.2.40 Importance Rating 4.7 Equipment Control: Ability to apply technical specifications for a system (EDGs).
Proposed Question: SRO Question # 90 PNPS is at rated conditions with an overhaul of the "B" EDG in progress. The plant is on day 3 of a 14 day LCO due to the diesel maintenance.
Then, the breaker for M0-1001-36B, RHR Loop B Torus Cooling Valve is determined to be inoperable and the valve cannot be opened from the control room.
M0-1001-34B, Torus Cooling/Spray Block Valve, is then verified closed and is disabled in the closed position.
All other equipment is operable.
If conditions do not change, the plant must be placed in COLD SHUTDOWN conditions within:
A. 1 day B. 4 days C. 8 days D. 12 days Proposed Answer: A Explanation (Optional):
A. Correct- The breaker for RHR valve 36B being INOP renders the "B" Torus Cooling and the "B" Containment Spray subsystems INOP. This in of itself is a 7 day LCO (TS 3.5.B.1 and 2). However the EDG outage requires that all low pressure core and containment cooling systems be operable lAW TS 3.5.F.1 while the EDG is not available. If this requirement is not met, the reactor must be placed in cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ( 1 day).
B. Incorrect- The plant must be in cold SO within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 4 days is obtained if the PNPS 2014 NRC Written 022014.doc
candidate applies the 151 part of LCO 3.5.F that addresses an LCO time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the diesel OOS and then adds 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to achieve Cold S/0.
C. Incorrect- The plant must be in cold SO within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 8 days is obtained if the candidate does not understand the relationship between the Torus Cooling and the Containment Spray subsystems and the lnop EDG. In this case the candidate would utilize the 7 day LCO of TS 3.5.B.1 and 2 and add 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to achieve cold SID.
D. Incorrect- The plant must be in cold SO within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 12 days is obtained if the candidate only considers M0-1001-36B as a PCIS valve. In that case then the TS action of disabling the M0-1001-34B would have restored the Primary Containment and the limiting LCO would then revert back to the original EDG LCO time. Since there is 11 days left on the original EDG LCO time and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to obtain cold SID the candidate could arrive at a time frame of 12 days.
Technical Reference(s): TS sections TS 3.5.B.1 and 2 and (Attach if not previously provided) section 3.5.F Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-09-06, EO 30 (As available)
Question Source: Bank # X Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 256000 A2.02 Importance Rating 2.9 A2.02- Ability to (a) predict the impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures Proposed Question: SRO Question# 91 The plant is at rated conditions with all components operable.
Then TBCCW non-essential block supply M0-4127 fails closed and cannot be re-opened.
Which one of the following is correct regarding:
(1) The impact on the Condensate system AND (2) The required action?
A. (1) Condensate pumps trip on low cooling water flow after 60 seconds.
(2) Enter PNPS 2.4.41, Loss of TBCCW, scram the reactor, secure all condensate and reactor feed pumps and close the MSIVs.
B. (1) Condensate pumps lose cooling water flow and begin to overheat.
(2) Enter PNPS 2.4.41, Loss of TBCCW, scram the reactor and secure all condensate and reactor feed pumps and close the MSIVs.
C. (1) Condensate pumps lose cooling water flow and begin to overheat.
(2) Enter PNPS 2.1.14, Station Power changes and insert the RPR array. Secure any condensate pump whose bearing temperatures exceed the limits of ARP C2R-B7, Omniguard Alarm.
D. (1) Condensate pumps trip on low cooling water flow after 60 seconds.
(2) Enter PNPS 2.4.49, Feedwater Malfunctions, scram the reactor, trip all reactor feed pumps when the condensate pumps trip. Close the MSIVs when vacuum can no longer be maintained.
Proposed Answer: B PNPS 2014 NRC Written 022014.doc
Explanation (Optional):
A. Incorrect: The condensate pumps will not automatically trip on low flow. Plausible in that the reactor feed pumps did until very recently.
B. Correct: The condensate pumps will lose cooling water flow. PNPS 2.4.41, Loss of TBCCW would be applicable procedure. The subsequent actions of this procedure are to scram, secure all feed and condensate and close the MSIVs.
C. Incorrect: The appropriate action would be to respond lAW 2.4.41. Plausible in that reducing power would allow removing a condensate pump from service and the referenced alarm does provide safe operating temperature limits for the condensate pumps.
D. Incorrect: The condensate pumps will not automatically trip. Also the MSIVs are closed when the last condensate pump is secured. Closing the MSIVs when vacuum begins to degrade is plausible because condensate flow cools the Gland Seal and SJAE condensers.
Technical Reference(s): PNPS 2.4.41, Subsequent action. (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-02-02-03 TO, RESPOND TO (As available)
LOSS OF TBCCW PUMPS.
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 202002 2.4.11 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of abnormal condition procedures. (Recirculation Flow Control)
Proposed Question: SRO Question # 92 The Reactor is operating at rated conditions when a FWLC malfunction results in a dual Recirc pump runback to minimum speed. Current plant conditions are:
- Reactor power: 78%
- Core flow: 33 MLbm/hr
- Both Recirc Pump scoop tubes have been locked Which one of the following is correct for the above conditions?
A. Enter PNPS 2.4.165, Reactor Core Instability, and if Decay Ratios are acceptable operation can continue at the current power level while establishing local control of the Recirc MG Sets.
B. lAW PNPS 2.1.14, Station Power Changes, control rods must be inserted in reverse order of the pull sheet to reduce power to <73%.
C. lAW PNPS 2.1.14, Station Power Changes, the RPR array must be inserted to reduce power to< 73%.
D. Enter PNPS 2.1.6, Reactor Scram, and insert a manual reactor scram.
Proposed Answer: C Explanation (Optional):
A. Incorrect: Power needs to be reduced. Plausible in that the reactor is also in the Buffer Zone. PNPS 2.4.165, subsequent action [5] directs that core decay ratios be checked against their limits. If acceptable, then operation can continue in the buffer zone.
However being above MELLA requires the power reduction.
B. Incorrect: Use of the RPR array is specified . Plausible in that if power were below the 60% load line, rods would be inserted using the reverse order of the pull sheet.
PNPS 2014 NRC Written 022014.doc
C. Correct: The plant is operating above the MELLA line. PNPS 2.4.20, subsequent action
[2] states: "IF Reactor power is above MELLLA line, THEN REDUCE power in accordance with PNPS 2.1.14 Section 7.9 until below the MELLLA line."
Lowering power to < 73% will restore operation to below the MELLA line. PNPS 2.1.14 section 7.9 directs use of the RPR since power is above the 60% load line.
D. Incorrect: The required action is to insert the RPR array. Plausible in that the plant is in the Operation Prohibited Region (above the MELLA line).
Technical Reference(s): PNPS 2.4.165, subsequent (Attach if not previously provided) actions PNPS 2.4.20, subsequent actions PNPS 2.1.14, section 7.9 Proposed References to be provided to applicants during examination: Power to Flow Map Learning Objective: 0-R0-03-03-09(02), TO RESPOND (As available)
TO A RECIRC SYSTEM SPEED OR FLOW CONTROL MALFUNCTION Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 204000 2.4.12 Importance Rating 4.3 Knowledge of general operating crew responsibilities during emergency operations. (RWCU)
Proposed Question: SRO Question # 93 The plant is at 30% power when the NRC contacts PNPS via the ENS line. The NRC reports that the FBI has determined that a credible threat exists specifically for Pilgrim station.
Details regarding the timing of the threat are not currently known.
Which one of the following is required?
A. Enter PNPS 2.1.14, Station Power Changes and commence a power reduction. No other action is required unless an attack is imminent.
B. Enter PNPS 2.1.14, Station Power Changes, commence a power reduction and isolate RWCU.
C. Enter PNPS 2.1.6 Reactor Scram, scram the reactor and isolate the Main Steam Lines.
D. Enter PNPS 2.1.6 Reactor Scram, scram the reactor and isolate RWCU.
Proposed Answer: D Explanation (Optional):
A. Incorrect: Because the threat is both credible and valid, a reactor scram is warranted.
Additionally isolating RWCU is required.
B. Incorrect: Because the threat is both credible and valid, a reactor scram is warranted.
C. Incorrect: The MSIVs are not isolated in order to preserve the main condenser as a heat sink.
D. Correct: The threat is credible because it came from the FBI. The threat is also classified as a Valid credible threat because it came from the NRC via the ENS line.
Attachment 1 of PNPS 5.3.14, Step [3] then requires a scram and that RWCU be isolated.
PNPS 2014 NRC Written 022014.doc
Technical Reference(s): PNPS 5.3.14, Attachment 1, (Attach if not previously provided) definitions and section 4.0.
Proposed References to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # PNPS 9453 Modified stem and distractors to include procedure.
Modified Bank#
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.1.25 Importance Rating 4.2 Ability to interpret reference materials, such as graphs, curves, tables, etc.
Proposed Question: SRO Question # 94 The plant is in cold shutdown with RHR aligned for Shutdown Cooling (SDC). Additional information is as follows:
- The plant was shutdown 3 days ago.
- Prior to the shutdown PNPS had been on line for 185 days.
- A loss of off-site power has just occurred.
- Immediately prior to the loss of off-site power, plant parameters were as follows:
o RPV level 30 inches o RPV moderator temperature 138 degrees o Drywell is open to allow repair of a leaking head vent valve Assuming no operator actions are taken, the 151 Tech Spec violation will occur within
_(1)_.
AND If a Tech Spec violation DOES occur, and the Shift Manager has determined that design isolation signals must be defeated in order to mitigate the event, NRC notification is required within_ (2) _ .
Notes: Do not consider any required NRC notifications required by the Emergency Plan.
References are provided.
A. (1) 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. (1) 6.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. (1) 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. (1) 6.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> PNPS 2014 NRC Written 022014.doc
Proposed Answer: A Explanation (Optional):
A. Correct: Per Attachment 5 of 2.4.25, and vessel level at 30", and 3 days after shutdown, heatup rate is 49.8 degrees/hr. A Tech Spec violation will occur when moderator
=
temperature reaches 212 degrees since the drywell is open. (212-138)/ 49.8 1.49 hrs.
As discussed in PNPS 2.4.25, page 9 and PNPS 1.3.12, Attachment 12, IF corrective action to recover from this situation involves inhibition of design isolation signals by jumpers/lifted leads/booted contacts or use of the Alternate Shutdown Panel, THEN notification to the NRC of the Station's intended actions is required. This is a 1-hour report as required by 10CFR50.54X.
B. Incorrect: The 151 Tech Spec Violation will occur when temperature exceeds 212 degrees. Plausible in 6. 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> is the amount of time that it will take for level to lower to TAF. If the candidate considers the cold SD safety limit of 12 inches above the TAF, the candidate may choose this response.
C. Incorrect: The NRC must be notified within 1hour. Plausible if the candidate does not recognize this as a 50.54X event and applies the reporting requirement for the initiation of a plant shutdown required by Tech Specs . (4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report by 50.72(b)(2)(i))
D. Incorrect: Incorrect the 1st tech spec violation will occur when moderator temperature reaches 212 degrees. Additionally, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report is required.
Technical Reference(s): PNPS 2.4.25, LOSS OF (Attach if not previously provided)
SHUTDOWN COOLING, Attachment 5 and page 9 PNPS 1.3.12, Attachment 2 Proposed References to be provided to applicants during examination: PNPS 2.4.25, LOSS OF SHUTDOWN COOLING, Attachment 5 PNPS 1.3.12, Attachment 2 Learning Objective: (As available)
Question Source: Bank #
Modified Bank # LOR Bank #55 Modified to include reporting requirements for 50.54.X PNPS 2014 NRC Written 022014.doc
New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.2.21 Importance Rating 4.1 Knowledge of pre- and post-maintenance operability requirements.
Proposed Question: SRO Question # 95 The plant is at rated conditions. HPCI has been removed from service for corrective maintenance. An LCO was generated lAW PNPS 1.3.34.2, Limiting Conditions for Operation Log.
Which of the following statements is/are correct regarding the closing of the LCO at the conclusion of the maintenance?
- 1 All condition reports documenting discrepancies identified during the maintenance must be closed.
- 2 Operability Evaluations must be completed for all condition reports documenting discrepancies identified during the maintenance.
- 3 Components that reposition or change state during an automatic initiation need not be in a Standby lineup provided all required power supplies are available to support an auto initiation.
A. 1 only B. 2 only C. 2 and 3 D. 1 and 3 Proposed Answer: B Explanation (Optional):
A. Incorrect: The condition report need not be closed. Only an Operability Eval must be completed.
PNPS 2014 NRC Written 022014.doc
B. Correct: lAW 1.3.34.2, section 6.5, step [2] (g), if the LCO condition involves a Condition Report documenting a degraded or nonconforming condition, then an approved Operability Evaluation must exist prior to the clearance of the LCO.
C. Incorrect: Statement 3 is not correct. Per PNPS 1.3.34.2 section 6.5 and Attachment 5, item [8], the system is required to be restored to a standby lineup or as determined by the SM to be acceptable for the current mode. Since the plant is at rated conditions, HPCI would be placed in a standby lineup.
D. Incorrect: The condition report need not be closed. Only an Operability Eval must be completed. Also HPCI is required to be in a standby lineup.
Technical Reference(s): PNPS 1.3.34.2 section 6.5 and (Attach if not previously provided)
Attachment 3 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01, 28a (As available)
Question Source: Bank #
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KJA# G2.3.13 Importance Rating 3.8 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.
Proposed Question: SRO Question# 96 A drywall entry is required. The Radiation Protection Manager has given approval for the entry.
Select the choice below that correctly completes the following regarding a containment entry lAW PNPS 1.4.12, Primary Containment Entry.
In order for the Shift Manager to authorize the containment entry WITHOUT obtaining any ADDITIONAL approvals:
- Reactor power level must be less than_ (1) _%.
- The containment atmosphere_ (2) _required to be de-inerted prior to the entry.
A. (1) 5 (2) is B. (1) 5 (2) is NOT
- c. (1) 10 (2) is D. (1) 10 (2) is NOT Proposed Answer: A Explanation (Optional):
A. Correct: When the containment is de-inerted and Reactor power level is less than 5%
CTP, Drywall entry may be authorized by the Shift Manager provided that approval has been obtained from the Radiation Protection Manager (RPM).
PNPS 2014 NRC Written 022014.doc
B. Incorrect: The containment must be de-inerted. Plausible in that there is a process for allowing entry into an inerted containment but that requires additional approvals.
C. Incorrect: Power must be< 5%. Plausible in that there is a process for allowing entry at higher powers but that requires additional approvals.
D. Incorrect: Power must be< 5% and the containment must be de-inerted. Plausible in that there is a process if these conditions cannot be met during emergencies but additional approvals are required.
Technical Reference(s): PNPS 1.4.12, section 4.1 and also (Attach if not previously provided)
Attachment 1 pages 2 and 3.
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-06-01 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.4.35 Importance Rating 4.0 Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects.
SRO Level Justification: lAW PNPS 2.1.7, VESSEL HEATUP AND COOLDOWN the CRS is required to review the OPER-07 data after each sheet of readings.
Proposed Question: SRO Question # 97 The plant was operating in single loop due to corrective maintenance on Recirc MG set "A".
The following then occurs:
- Recirc MG Set "B" trips
- The reactor is manually scrammed
- A reactor plant cooldown is commenced using the main turbine bypass valves Additional information is as follows:
- RPV level is +30 inches on the Narrow Range indicators
- OPER-07 Cooldown Data is as shown in the table below
- At time 00:60, the fault on Recirc MG Set "B" is repaired Date TR-260-151A & B PR/FR-640-27 TR-263-104 TR-263-104 PaneiC904 Panel C905 Panel C921 Panel C921 (Pt. 2) (Pt. 2) (Pt. 4) (Pt. 9)
Time Recirc/RHR Loop RxWide Vessel Line Vessel Below Vessel Suction Temp Range Shell Adj Vessel Water Bottom Loop A Loop B Pressure To Flange Drain Level Head (RXX022) (RXX024) (RXX024) (RXX030) 00:00 510 500 900 531 505 528 506 00:15 499 492 800 518 468 516 467 00:30 472 468 700 503 428 500 .430 00:45 450 448 600 486 365 485 370 00:60 425 426 500 467 298 464 308 lAW PNPS 2.4.24, Reactor Vessel Cold Water Stratification, and the information provided, which one of the following is actions is authorized?
Directing the ...
PNPS 2014 NRC Written 022014.doc
A. RO to start Recirc MG Set "B".
B. RO to raise RPV level to > + 60 inches as indicated on Ll-263-1 01.
C. NLO to increase RWCU system flow to the filter demineralizer stretch rating.
D. NLO to throttle close M0-1201-85, RWCU Recirculation Loop Suction Valve.
Proposed Answer: D Explanation (Optional):
A. Incorrect: There is a greater than 145 oF between the bottom head drain temperature and saturation temperature. Tech Spec 3.6.A, and OPER 19 preclude starting the recirc pump. Plausible in that this would be an acceptable action to reduce vessel stratification if recirc pump start limitations were met.
B. Incorrect: Although the vessel is stratifying, RPV level is not raised because the main condenser is being used as a heat sink. This action is only taken if a heat sink is not available in order to increase natural circulation.
C. Incorrect: This action is not directed by PNPS 2.4.24. Plausible in the procedure does direct increasing RWCU flow from the bottom head drain.
D. Correct: The vessel is stratifying. Throttling closed on the RWCU-85 valve will result in increasing flow from the bottom head region.
Technical Reference(s): PNPS 2.4.24, Attachment 1, sheet (Attach if not previously provided) 1.
Tech Spec 3.6.A and OPER 19 start limitations.
Proposed References to be provided to applicants during examination: Steam Tables Learning Objective: 0-R0-02-06-05 , EO 14a (As available)
Question Source: Bank #
Modified Bank #
New X PNPS 2014 NRC Written 022014.doc
Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.2.23 Importance Rating 4.6 Ability to track Technical Specification limiting conditions for operations.
KIA Justification: In order to track Tech Spec LCOs it is necessary to know when the clock started for the LCO.
Proposed Question: SRO Question# 98 Given the following sequence:
- At time 00:00: RBCCW Pump "A" is started and pump "8" secured
- At time 00:30: An operator reports that the RBCCW Pump "A" seal is leaking
- At time 01 :00: The seal leak is quantified as 40 drops/minute
- At time 01 :20: The Shift Manger contacts the system engineer to assist in an Operability Determination
- At time 03:00: The system engineer reports that the total RBCCW loop "A" leakage now exceeds the 240 drops per minute limit and that the RBCCW loop will not meet its required mission time of 30 days.
- At time 03:15: The Shift Manager reviews and concurs with the system engineer's calculations and declares the "A" RBCCW loop inoperable.
In accordance with EN-OP-104, Operability Determination Process, the time clock (LCO time) for the associated Limiting Condition of Operation, Action Statement started at:
A. 00:00 since the seal most likely began leaking upon initial start.
B. 00:30 since this was the time of discovery.
C. 01 :00 since this is the time the leakage was quantified and when the limit of 240 gpm was known to have been exceeded.
D. 03:15 since this was the time the shift manager declared the loop inoperable.
Proposed Answer: D Explanation (Optional):
A. Incorrect: Per EN-OP-104, the clock time starts when the SM makes the operability determination.
PNPS 2014 NRC Written 022014.doc
B. Incorrect: Per EN-OP-104, the clock time starts when the SM makes the operability determination.
C. Incorrect: Per EN-OP-104, the clock time starts when the SM makes the operability determination.
D. Correct: Per EN-OP-104, and regarding Operability Determination Process Times the Time of Determination is the moment the Shift Manager (SM) decides on the status of the condition. The time clock for a Limiting Condition of Operation, Action Statement or reportable condition starts at the time of determination.
Technical Reference(s): EN-OP-104, Operability (Attach if not previously provided)
Determination Process, page 15 Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-06-01-06, TO, 299-03-05-005 (As available)
CHECK SYSTEMS GOVERNED BY TECH SPECS. TO ENSURE OPERABILITY REQUIREMENTS ARE MET.
Question Source: Bank#
Modified Bank# PNPS 11479 Modified to provide an actual example New Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.3.11 Importance Rating 4.3 Ability to control radiation releases.
Proposed Question: SRO Question # 99 A main steam leak in the turbine building has occurred that cannot be isolated. EOP-05, Rad Release Control, has been entered.
lAW EOP-05, an Emergency Depressurization is required before ...
An EAL chart is provided for your use.
A. Main Stack High Range Effluent Monitor Rl-1 001 has exceeded 2 R!Hr for 15 minutes.
B. Main Stack Low Range PRMs 1705-18A and B have exceeded 400,000 cps for 15 minutes.
C. The dose at the site boundary exceeds 100 mRem whole body as indicated by dose assessment.
D. The dose at the site boundary exceeds 1000 mRem whole body as indicated by dose assessment.
Proposed Answer: D Explanation (Optional):
A. Incorrect: EOP-05 requires that an Emergency Depressurization be conducted before the General Emergency Offsite Release limits are exceeded. A Main Stack High Range monitor of 2 R!Hr for 15 minutes is the threshold for a Site Area Emergency.
B. Incorrect: Main Stack Low Range PRMs reading 400,000 cps for 15 minutes is the threshold for an Alert. Plausible in that there is a specified EOP-05 action to be taken when the offsite release exceeds the limits of an Alert EAL.
C. Incorrect: A dose assessment that indicates that the dose rate at the site boundary will exceed 100 mRem/hr is the threshold for a Site Area Emergency.
D. Correct: EOP-05 requires that an Emergency Depressurization be conducted before the General Emergency Offsite Release limits are exceeded. A General Emergency PNPS 2014 NRC Written 022014.doc
EAL has been exceeded when dose projections indicate that the dose rate at the site boundary will exceed 1000 mRem/hr.
Technical Reference(s): EOP-05, step (Attach if not previously provided)
EAL Chart Proposed References to be provided to applicants during examination: HOT EAL Chart Learning Objective: 0-R0-03-04-07, EO ?a (As available)
Question Source: Bank #
Modified Bank #
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
Examination Outline Cross-reference: Level RO SRO Tier# 3 Group#
KIA# G2.4.34 Importance Rating 4.1 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.
KIA Justification: lAW PNPS 2.4.143, operation of the SRVs from their ASPs can only be assigned to licensed ROs.
Proposed Question: SRO Question # 100 The Main Control Room has just been abandoned due to a fire.
What action is required to be directed regarding the Alternate Shutdown Panels for the SRVs?
A Place all SRVs in Close at their ASPs within 15 minutes to terminate any spurious SRV operation.
B. Place all SRVs in Close at their ASPs within 24 minutes to terminate any spurious SRV operation.
C. Take local control of at least one SRV within 24 minutes and control pressure between 900 and 1050 psig to prevent excessive SRV cycling on high pressure.
D. Take local control of at least one SRV within 15 minutes and control pressure between 900 and 1050 psig to prevent excessive SRV cycling on high pressure.
Proposed Answer: A Explanation (Optional):
A Correct: lAW the Caution on page 9 of PNPS 2.4.143, local control of all SRVs must be established within 15 minutes to prevent spurious SRV operation.
B. Incorrect: The required time frame is 15 minutes. The 24 minute time frame is the time frame for establishing injection flow.
C. Incorrect: There is no directed action to establish pressure control within a specified time frame. However this is the specified band when pressure control is first established.
PNPS 2014 NRC Written 022014.doc
D. Incorrect: There is no directed action to establish pressure control within a specified time frame.
Technical Reference(s): 2.4.143 page 9 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: None Learning Objective: 0-R0-03-03-26(01) TO, SUPERVISE (As available)
CREW DURING SHUTDOWN OUTSIDE THE CONTROL ROOM Question Source: Bank#
Modified Bank#
New X Question History: Last NRC Exam: Not used Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Comments:
PNPS 2014 NRC Written 022014.doc
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